Qinshan NPP in-core fuel management improvement

International Nuclear Information System (INIS)

Kong Deping; Liao Zejun; Wu Xifeng; Wei Wenbin; Wang Yongming; Li Hua

2006-01-01

In the 10-year operation of Qinshan Nuclear Power Plant, the initial designed reloading strategy has been improved step by step based on the operation experiences and the advanced domestic and international fuel management methods. Higher burnup has been achieved and more economic operation gained through the loading pattern improvement and the fuel enrichment increased. The article introduces the in-core fuel management strategy improvement of Qinshan Nuclear Power Plant in its 10-year operation. (authors)

Testing and operation of nuclear air-cleaning systems in Qinshan NPP

International Nuclear Information System (INIS)

Yang Lin

1993-01-01

The components of nuclear air-cleaning system, system function, operational mode and the performance of cleaning components in Qinshan Nuclear Power Plant are described. The items, purpose, methods and results of in-place testing after the installation are also described in detail. The in-place testing verifies that nuclear air-cleaning systems in Qinshan Nuclear Power Plant are reliable and high effective. It also describes the points of the operational management. It is shown that the good operational management is the key which developed prescription function of nuclear air-cleaning systems. At present, Qinshan Nuclear Power Plant will be in full power, the normal operation of the system is satisfied with the demand of safe operation in Qinshan Nuclear Power Company

Successful completion of the Qinshan phase III nuclear power plant-a successful model for Chinese-Canadian cooperation

International Nuclear Information System (INIS)

Peng Xiaoxing

2004-01-01

This report documents Qinshan CANDU project construction and commissioning experience as well as management strategies and approaches that contributed to the successful completion of the project. The Qinshan phase III (CANDU) nuclear power plant was built in record times: Unit 1 achieved commercial operation on December 31, 2002 and Unit 2 on July 24, 2003, 43 days and 112 days ahead of schedule respectively. The reference plant design is the Wolsong 3 and 4 CANDU-6 units in the Republic of Korea. Improvements in design and construction methods allowed Unit 1 to be constructed in 51.5 Months from First Concrete to Criticality-a record in China for nuclear power plants. The key factors are project management and project management tools, quality assurance, construction methods (including open top construction, heavy lifts and modularization), electronic documentation with configuration control that provides up-to-date on-line information, CADDS design linked with material management, specialized material control including bar coding, and planning. The introduction of new design and construction techniques was achieved by combining conventional AECL practices with working experiences in China. The most advanced tools and techniques for achieving optimum construction quality, schedule and cost were used. Successful application of advanced project management methods and tools will benefit TQNPC in operation of the station, and the Chinese contractors in advancing their capabilities in future nuclear projects in China and enhancing their opportunities internationally. TQNPC's participation in Quality surveillance (QS) activities of nuclear steam plant (NSP) and Balance of Plant (BOP) offshore equipment benefited TQNPC in acquiring knowledge of specific equipment manufacturing processes, which can be applied to similar activities in China. China has established the capability of manufacturing CANDU fuel and becoming self-reliant in fuel supply. Excellent co-operation and

The modification of main steam safety valves in Qinshan phase Ⅱ expansion project

International Nuclear Information System (INIS)

Chen Haiqiao

2012-01-01

The main steam safety valves of NPP steam system are second- class nuclear safety component. It used to limit the pressure of SG secondary side and main steam system via emitting steam into the environment. At present, the main steam safety valves have mechanical valves and assisted power valves. According to the experience of power plants at home and abroad, including Qinshan Phase Ⅱ unit 1/2 experience feedback, Qinshan Phase Ⅱ expansion project made modification on valve type, setting value and valve body. This paper introduce the characteristics of different safety valve types, the modification of main steam safety valves and the modification analysis on safety issues.security and impact on the other systems in Qinshan Phase Ⅱ expansion project. (author)

Qinshan 300Mwe NPP full scope simulator upgrade

International Nuclear Information System (INIS)

Qi Kelin; Li Qing; Liu Wei, Lai Shengyuan

2006-01-01

On April 28,2004, RINPO was awarded the project for Qinshan 300Mwe NPP full scope simulator upgrade, the SAT (site acceptance test) was completed on June 30 2005 and the simulator put into operator training again. Scope of upgrade includes: computer system (DGI server and workstations) all replaced by microcomputers; G2 I/O controllers all replaced by RTP EIOBC; Unix-based simulation support environment replaced by RINPO's PC-based simulation environment RINSIMTM, Instructor software replaced by RINPO's PC-based instructor software with function and diagram redesigned; DEH, Feed-water control and some other digital control systems redeveloped to follow NPP modifications; desk-top simulator with soft panel control room developed as byproduct; most of the models not changed but it is planned the reactor core and PPC model will be upgraded in near future. SAT of upgrade demonstrates that the performance of the simulator much improved after the upgrade. (author)

Upgrading Planning and Executive Strategy for Reactor Protection System and Relative Equipment in Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Jiang Zuyue

2010-01-01

Qinshan Nuclear Power Plant (QNPP) is the first nuclear power plant in China which completed the reactor protection system (RPS) upgrading with new digital safety instrumentation and control (I and C) platform instead of original analog system. At the same time,the nuclear instrumentation system (NIS) was upgraded with the same digital I and C platform. For adapting QNPP's actual engineering situation,the upgrading planning was taken by comprehensively investigating current development and application of digital safety I and C platform in the worldwide scope and by reviewing plant's original systems operation history. The project executive strategy-QNPP's leading role with necessary overseas cooperation and internal technical supports as great as possible, was determined. Some significant factors might influence and restrict the RPS and relative equipment upgrading executive actions in an operating NPP were analyzed.Finally, the engineering feasibility was briefly assessed to recognize the anticipated issues and difficulties and to prepare the relative solutions in advance for the purpose of ensuring the RPS upgrading objectives completely realized. (authors)

Activity level of gross α and gross β in airborne aerosol samples around the Qinshan NPP

International Nuclear Information System (INIS)

Chen Bin; Ye Jida; Chen Qianyuan; Wu Xiaofei; Song Weili; Wang Hongfeng

2007-01-01

The monitoring results of gross α and gross 13 activity from 2001 to 2005 for environmental airborne aerosol samples around the Qinshan NPP base are presented in this paper. A total of 170 aerosol samples were collected from monitoring sites of Caichenmen village, Qinlian village, Xiajiawan village and Yangliucun village around the Qinshan NPP base. The measured specific activity of gross α and gross β are in the range of 0.02-0.38 mBq/m 3 and 0.10-1.81 mBq/m 3 , respectively, with an average of 0.11 mBq/m 3 and 0.45mBq/m 3 , respectively. They are lower than the average of 0.15 mBq/m 3 and 0.52 mBq/m 3 , of reference site at Hangzhou City. It is indicated that the specific activity of gross α and gross β for environmental aerosol samples around the Qinshan NPP base had not been increased in normal operating conditions of the NPP. (authors)

Safety design of Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Ouyang Yu; Zhang Lian; Du Shenghua; Zhao Jiayu

1984-01-01

Safety issues have been greatly emphasized through the design of the Qinshan Nuclear Power Plant. Reasonable safety margine has been taken into account in the plant design parameters, the design incorporated various safeguard systems, such as engineering safety feature systems, safety protection systems and the features to resist natural catastrophes, e. g. earthquake, hurricanes, tide and so on. Preliminary safety analysis and environmental effect assessment have been done and anti-accident provisions and emergency policy were carefully considered. Qinshan Nuclear Power Plant safety related systems are designed in accordance with the common international standards established in the late 70's, as well as the existing engineering standard of China

Present status and future development of Qinshan nuclear power project

International Nuclear Information System (INIS)

Ouyang Yu

1987-01-01

Qinshan 300 MWe Nuclear power Project is the first domestically designed and constructed nuclear power plant in China. Here is a brief description of its progress in design work, equipment manufacture and site construction since the first structural concrete in March 1985. In Qinshan area four units of 600 MWe each are planned to be built with collaboration of proper foreign partners. (author)

Ultrasonic measurement on RPV stud-bolt loading under hot transient of Qinshan NPP

International Nuclear Information System (INIS)

Qu Jiadi; Dou Yikang; Zhu Shiming; Lu Jie; Wang Yingguan

1994-01-01

It is a continuation of research work for sealing analysis and tests on the PRV of PWR. It expounds that the key of solving thermal transient sealing problem lies in giving the thermal increment of stud-bolt fatigue life and transient loading spectrum for vessel analysis. The authors recounted the fundamental works and main results of ultrasonic measurement on RPV stud-bolt loading on the reactor of Qinshan Nuclear Power Plant. The measuring capability exceeds 1 m length and 300 degree C temperature. Therefore, it is possible to be used in the field of NPP

The application of plant information system on third Qinshan nuclear power plant

International Nuclear Information System (INIS)

Liu Wangtian

2005-01-01

Plant overall control has been applied in Qinshan Nuclear Power Plant, which enhances the security of plant operation, but it is not enough to improve the technical administration level. In order to integrate the overall information and to improve the technical administration level more. Third Qinshan Nuclear Power Plant applies the plant information system. This thesis introduces the application of plant information system in Third Qinshan Nuclear Power Plant and the effect to the plant after the system is carried into execution, in addition, it does more analysis and exceptions for application of plant information system in the future. (authors)

Structure study and design of Qinshan NPP PCCV

International Nuclear Information System (INIS)

Xia Zufeng; Xu Yongzhi; Wang Tianzhen; Wu Jibiao

1993-02-01

The design process of Qinshan NPP (nuclear power plant) PCCV (prestressed concrete containment vessel) is summarized. The tendon test, structural description, design bases and analysis method are introduced. The arrangement for preventing concrete from cracking and design features of post-tensioning system and steel liner are presented. The results of model test and non-linear analysis for ultimate load in Qinshan NPP PCCV are also given. Through the integrity test of PCCV, it shows that the test values are in agreement with predicted values, the structure is excellent and the performance of leak tightness conforms to the safety requirements

Self-reliance and innovation of Qinshan phase II NPP project

International Nuclear Information System (INIS)

Ye Qizhen; Yang Lanhe

2007-01-01

This article mainly describes the self-reliance and innovation of Qinshan nuclear power project of phase II, in-between it contains new reactor core design, as well as related experimental and calculation analysis, especially for new reactor design produced fluid-induced vibration model test, theoretical analysis and testing in-built reactor; aiming at two-loop NSSS a series improvement made for safety systems and related safety analysis to enhance their reliability and redundancy; according to specialty of two-loop NSSS an optimization made for NPP parameters and design of related equipments, for the purpose to make the output of NPP maximal; design of main reactor building and T-G building also improved according to characteristics of two-loop NSSS and site conditions. CRDM and refueling machine are researched and manufactured on base of self-reliance, their performance are better than design requirements, large portion of key equipments are localized through different way. In construction first time realized the integrated erection of containment dome. During the commissioning non-nuclear steam driving of T-G set, as well as 500 kV high voltage rising using emergent diesel generator, etc. are carried out.In period of operation still continuous innovation and improvement are made, so that to keep the good record of operation. (authors)

Fracture mechanics analysis and evaluation for the RPV of the Chinese Qinshan 300 MW NPP and PTS

International Nuclear Information System (INIS)

He Yinbiao; Isozaki, Toshikuni

2000-03-01

One of the most severe accident conditions of a reactor pressure vessel (RPV) in service is the loss of coolant accident (LOCA). Cold safety injection water is pumped into the downcomer of the RPV through inlet nozzles, while the internal pressure may remain at high level. Such an accident is called pressurized thermal shock (PTS) transient according to 10 CFR 50.61 definition. This paper illustrates the fracture mechanics analysis for the existing RPV of the Chinese Qinshan 300 MW nuclear power plant (NPP) under the postulated PTS transients that include SB-LOCA, LB-LOCA of Qinshan NPP and Rancho Seco transients. 3-D models with the flaw depth range a/w=0.05∼0.9 (a: flaw depth; w: wall thickness) were used to probe what kind of flaw and what kind of transient are most dangerous for the RPV in the end of life (EOF). Both the elastic and elastic-plastic material models were used in the stress analysis and fracture mechanics analysis. The different types of flaw and the influence of the stainless steel cladding on the fracture analysis were investigated for different PTS transients. comparing with the material initiation crack toughness K IC , the fracture evaluation for the RPV in question under PTS transients are performed in this paper. (author)

Qinshan CANDU commissioning - a successful partnership

International Nuclear Information System (INIS)

Alikhan, S.; Thomson, J.; Jun, G.; Guoyuan, J.

2004-01-01

The Qinshan CANDU Nuclear Power Plant consists of 2 x 728 MWe CANDU 6 units, built in Zhejiang Province, China, by the Third Qinshan Nuclear Power Company (TQNPC) as the owner and Atomic Energy of Canada Limited (AECL) as the main contractor. The Contract between China National Nuclear Corporation (CNNC) and AECL was signed in November 1996 and became effective on February 12, 1997 with scheduled completion dates of February 12, 2003 for Unit 1 and November 12, 2003 for Unit 2. Unit 1 was declared in-service on December 31, 2002, 43 days ahead of schedule and Unit 2 was declared in service on July 20, 2003, 115 days ahead of schedule. The successful partnership between AECL, Bechtel, Hitachi and TQNPC working as a team is the key to this success. Total commissioning period from first energization of the system service transformer to in-service for both units was 20.7 months, which is significantly better than the experience at other comparable CANDU 6 units. It has clearly demonstrated the benefits of building two units together, about 6 months apart, to achieve optimum utilization of resources already mobilized for the first unit; the second unit is commissioned with less than 40% of the effort required for the first unit. Since in-service to the end of March 2004, Unit 1 has operated at a gross capacity factor of 93% and Unit 2 at 82.5%, including loss of production for one month in August 2003 to repair the failure of turbine LP blades tie-wire. (author)

Thermal-hydraulic calculation and analysis for QNPP (Qinshan Nuclear Power Plant) containment

International Nuclear Information System (INIS)

Xie Hui; Zhou Jie; He Yingchao

1993-01-01

Three containment thermal-hydraulic codes CONTEMPT-LT/028, CONTEMPT-4/MOD3 and COMPARE are used to compute and analyse the Qinshan Nuclear Power Plant (QNPP) containment response under LOCA or MSLB conditions. An evaluation of the capability of containment of QNPP is given

Concentration and distribution of 14C in aquatic environment around Qinshan nuclear power plant

International Nuclear Information System (INIS)

Wang Zhongtang; Guo Qiuju; Hu Dan; Xu Hong

2015-01-01

In order to study the concentration and distribution of 14 C in aquatic environment in the vicinity of Qinshan Nuclear Power Plant (NPP) after twenty years' operation, an apparatus extracting dissolved inorganic carbon from water was set up and applied to pretreat the water samples collected around Qinshan NPP. The 14 C concentration was measured by accelerator mass spectrometer (AMS). The results show that the 14 C specific activities in surface seawater samples range from 196.8 to 206.5 Bq/kg 203.4 ± 5.6) Bq/kg in average), which are close to the background. The 14 C concentrations in cooling water discharged from Qinshan NPP are close to the 14 C values in near shore seawater samples out of liquid radioactive effluent discharge period. It can be further concluded that the 14 C discharged previously is diluted and diffused well, and no 14 C enrichment in seawater is found. Also, no obvious increment in the 14 C specific activities of surface water and underground water samples are found between Qinshan NPP region and the reference region. (authors)

Qinshan Phase III (CANDU) nuclear power project quality assurance

International Nuclear Information System (INIS)

Wang Lingen; Du Jinxiang

2001-01-01

The completion and implementation of quality assurance system of Qinshan Phase III (CANDU) nuclear power project are presented. Some comments and understanding with consideration of the project characteristics are put forward

The Qinshan phase III project-a successful model of sino-canadian cooperation

International Nuclear Information System (INIS)

Pang, S.H.H.; Alikhan, S.; Gu Jun

2005-01-01

The Qinshan Phase III (CANDU) Project, the largest-scale cooperative project between China and Canada, was completed in 2003 well in advance of the schedule and 10% under budget. The Third Qinshan (Phase III) Nuclear Power Plant (TQNPP) was built in record times: Unit 1 achieved commercial operation on December 31, 2002 and Unit 2 on July 20, 2003, 43 days and 115 days ahead of schedule respectively. Improvements in design and construction methods allowed Unit 1 to be constructed in 51.5 months from First Concrete to Criticality - a record in China for nuclear power plants. The key factors are project management and project management tools, quality assurance, construction methods, electronic documentation with configuration control that provides up-to-date on-line information, CADDS design linked with material management and control. New design and construction techniques were introduced by combining conventional AECL practices with working experiences in China. The most advanced tools and techniques for achieving optimum construction quality, schedule and cost were used. Successful application of advanced project management methods and tools has benefited TQNPC in its subsequent plant operation, and the Chinese contractors in advancing their capabilities in future nuclear projects in China as well as enhancing their opportunities internationally. Excellent co-operation and teamwork within the integrated TQNPC/AECL Commissioning Team with well documented QA program, process and procedures also contributed to the remarkable success of the Project. AECL's initial assessment, based on lessons learned, showed that the project schedule could readily be reduced to 66 months and the capital costs reduced by 25% for a replication project. AECL is building on this experience and successful results of TQNPP in its Advanced CANDU Reactor TM (ACR TM ) ** design. (authors)

Brief account of the design philosophy for third Qinshan NPP shutdown safety system based on practical application

International Nuclear Information System (INIS)

Xiong Weihua

2005-01-01

Qinshan CANDU power plant is uses the Canadian proven CANDU6 nuclear power technology. It has two characteristic: 1. heavy water-as moderator and coolant; 2. natural uranium as the fuel and change fuel during normal operating. CANDU6 include four special safety system: the No.1 shutdown system (SDS No.1), the No.2 shutdown system (SDS No.2), the containment system, the emergency core cooling system (ECCS). QinShan CANDU power plant is the first commercial PHWR nuclear power plant in China. And some aspect is not similar to everybody. The intention of the article is to introduce the basic design and functions. (authors)

Computer-aided engineering for Qinshan CANDU projects

International Nuclear Information System (INIS)

Huang Zhizhang; Goland, D.

1999-01-01

The author briefly describes AECL's work in applying computer-aided engineering tools to the Qinshan CANDU Project. The main emphases will be to introduce the major CADD software tools and their use in civil design, process design and EI and C design. Other special software tools and non-CADD tools and their applications are also briefly introduced

Design and test of the borosilicate glass burnable poison rod for Qinshan nuclear power plant core

International Nuclear Information System (INIS)

Huang Jinhua; Sun Hanhong

1988-08-01

Material for the burnable poison of Qinshan Nuclear Power Plant core is GG-17 borosilicate glass. The chemical composition and physico-chemical properties of GG-17 is very close to Pyrex-7740 glass used by Westinghouse. It is expected from the results of the experiments that the borosilicate glass burnable poison rod can be successfully used in Qinshan Nuclear Power Plant due to good physical, mechanical, corrosion-resistant and irradiaton properties for both GG-17 glass and cold-worked stainless steel cladding. Change of material for burnable poison from boron-bearing stainless steel to borosilicate glass will bring about much more economic benefit to Qinshan Naclear Power Plant

Test research and analysis for ultimate capacity of Qinshan NPP PCCV

International Nuclear Information System (INIS)

Zufeng, X.

1994-01-01

This paper introduces design and research for containment of Qinshan NPP which is the first PWR in CHINA designed and constructed by ourselves. The PCCV design is basically in conformity to ASME code. To verify the structural integrity capacity of Qinshan NPP containment, we fulfilled SIT and ILRT successfully in June, 1991. The special attention of the paper is focused on the ultimate capacity of the PCCV under severe accidents and earthquake. A study comprised of five different independent parts has been performed for the development of containment model test and corresponding nonlinear analysis. There are two prestressed concrete containment models with equipment hatch. One is 1/15 scale with steel liner tested on shake table and then moved out loaded with atmospheric pressure. The other is 1/10 scale without steel liner loaded with water pressure until destruction. From different methods including model test and nonlinear analysis, all obtained unanimous conclusion. The capacity under internal pressure and earthquake is reliable. The safety margin is enough. Consequently, in the second phase of Qinshan NPP and other PWR NPP under design, PCCV should be a better selection in China since it's more economic, rational and safe. (author)

Nuclear power reactor safety research activities in CIAE

International Nuclear Information System (INIS)

Pu Shendi; Huang Yucai; Xu Hanming; Zhang Zhongyue

1994-01-01

The power reactor safety research activities in CIAE are briefly reviewed. The research work performed in 1980's and 1990's is mainly emphasised, which is closely related to the design, construction and licensing review of Qinshan Nuclear Power Plant and the safety review of Guangdong Nuclear Power Station. Major achievements in the area of thermohydraulics, nuclear fuel, probabilistic safety assessment and severe accident researches are summarized. The foreseeable research plan for the near future, relating to the design and construction of 600 MWe PWR NPP at Qinshan Site (phase II development) is outlined

Safety assessment to support NUE fuel full core implementation in CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Fan, H.Z.; Laurie, T.; Siddiqi, A.; Li, Z.P.; Rouben, D.; Zhu, W.; Lau, V.; Cottrell, C.M. [CANDU Energy Inc., Mississauga, Ontario (Canada)

2013-07-01

The Natural Uranium Equivalent (NUE) fuel contains a combination of recycled uranium and depleted uranium, in such a manner that the resulting mixture is similar to the natural uranium currently used in CANDU® reactors. Based on successful preliminary results of 24 bundles of NUE fuel demonstration irradiation in Qinshan CANDU 6 Unit 1, the NUE full core implementation program has been developed in cooperation with the Third Qinshan Nuclear Power Company and Candu Energy Inc, which has recently received Chinese government policy and funding support from their National-Level Energy Innovation program. This paper presents the safety assessment results to technically support NUE fuel full core implementation in CANDU reactors. (author)

Severe accident management guidance for third Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Su Changsong

2010-01-01

The paper describes the background, document structure and the summaries of Severe Accident Management Guidance (SAMG) for Third Qinshan Nuclear Power Plant (TQNPP), and also introduces briefly some design features and their impacts on SAMG. (authors)

Two codes used in analysis of rod ejection accident for Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Zhu Xinguan

1987-12-01

Two codes were developed to analyse rod ejection accident for Qinshan Nuclear Power Plant. One was based on point model with temperature reactivity feedback. In this code, the worth of ejected rod was obtained under'adiabatic' approximation. In the other code, the Nodal Green's Function Method was used to solve space-time dependent neutron diffusion equation. Using these codes, the transient core-power have been calculated for two rod ejection cases at beginning of core-life in Qinshan Nuclear Power Plant

Coupling analysis on the soft ground settlement laws in Qinshan nuclear power phase I sea wall project

International Nuclear Information System (INIS)

Sun Feng; Pan Rong; Zhu Xiuyun; Zhang Dingli

2011-01-01

Qinshan Nuclear Power Phase I sea wall project is a barrier engineering in defending the design basis flooding, which is of importance to the safety of NPP. The geological condition has the feature of high compressibility and low penetration, such as the soft ground of 1 + 450 section of Qinshan Nuclear Power Phase I sea wall. Based on parameters acquired from the site experiment, 3-D finite difference analysis is put forward to study the feature of consolidation settlement laws, which can embody the fluid-solid coupling interaction. The conclusions of numerical analysis agree well with the in-site measured data, and it, can contribute to the design and construction of raising sea wall project. (authors)

New experience on construction and installation work in Qinshan PHWR nuclear power plant

International Nuclear Information System (INIS)

Lu Huaxiang

2004-01-01

The article provides a summary of the new experience on construction management and construction technology in the field of civil construction and installation work in Qinshan PHWR nuclear power plant, with focus on innovation in project management mode, new technology application and computerized management of construction and installation work. Management innovation, technical innovation and information technology are the key contributors to overall success of Qinshan PHWR nuclear power plant in construction and installation work. The new experience derived in these fields will be of great significance to promote independent construction of the new-round nuclear power projects in China. (author)

Analysis research on mixing characteristics of lower plenum of Qinshan phase Ⅱ NPP by CFD method

International Nuclear Information System (INIS)

Mao Huihui; He Peifeng; Lu Chuan; Zhang Hongliang

2015-01-01

The flowing and mixing characteristics of the lower plenum of Qinshan Phase n NPP were analyzed by CFD method. The calculation results were compared with the results of the reactor hydraulic simulation test. On core inlet mass flow distributions, both upwind and high resolution advection schemes show good agreements with test results. While on lower plenum mixing characteristics, the calculation results from either upwind or high resolution advection schemes show relatively large differences to the test data. Relatively, upwind advection schemes predict better anticipations on maximum and minimum mixing factors. Furthermore, whether or not considering helix flow by main pump is the most possible key factor that leads to difference between CFD calculation and test results. (authors)

Modelling and simulation of containment on full scope simulator for Qinshan 300 MW Nuclear Power Unit

International Nuclear Information System (INIS)

Zou Tingyun

1996-01-01

A multi-node containment thermal-hydraulic model has been developed and adapted in Full Scope Simulator for Qinshan 300 MW Nuclear Power Unit with good realtime simulation effects. Containment pressure for LBLOCA calculated by the model is well agreed with those of CONTEMPT-4/MOD3

Qualification test of chemical cleaning for secondary side of steam generator in Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Zhang Mengqin; Zhang Shufeng; Yu Jinghua; Hou Shufeng

1997-07-01

The chemical cleaning technique for removing sludge on the secondary side in Qinshan Nuclear Power Plant has been qualified. The chemical cleaning process will carry out during shutdown refuelling. The qualification test has studied the effect of chemical cleaning agent component, cleaning time on dissolution effectiveness of sludge (Fe 3 O 4 ) and to evaluate corrosion situation of main materials of SG in the cleaning process. The main component of cleaning agent is EDTA. The cleaning temperature is 20∼30 degree C. It is determined that allowable remains amount of cleaning agent (EDTA). The technique of cleaning, rinse, passivation for the chemical cleaning in Qinshan Nuclear Power Plant has been made. The qualification test shown that the technique can dissolve Fe 3 O 4 >1 g/L, the corrosion of materials is in allowable value, the allowable remains of EDTA is <0.01%. The technique character is static, ambient temperature. (9 refs., 12 tabs.)

Improved design on Qinshan 300 MWe nuclear power plant

International Nuclear Information System (INIS)

Shi Peihua; Cheng Wanli; Lu Rongliang

1993-01-01

The main aim, guiding ideology, general performance and parameters of improved design on Qinshan 300 MWe nuclear power plant are presented. Improved items are also introduced including the characteristic of layout in nuclear island building, decreasing unnecessary devices increasing necessary safety facilities and unifying code and standard. The progress of improved design is presented

Improved design on Qinshan 300 MWe nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Peihua, Shi; Wanli, Cheng; Rongliang, Lu [Shanghai Nuclear Engineering Research and Design Inst. (China)

1993-06-01

The main aim, guiding ideology, general performance and parameters of improved design on Qinshan 300 MWe nuclear power plant are presented. Improved items are also introduced including the characteristic of layout in nuclear island building, decreasing unnecessary devices increasing necessary safety facilities and unifying code and standard. The progress of improved design is presented.

Lessons learned from current Qinshan CANDU project and the impact on future NPP's

International Nuclear Information System (INIS)

Hedges, K. R.; Didsbury, R.; Yu, S. K. W.

2000-01-01

AECL has adopted an evolutionary approach to the development of the CANDU 6 and CANDU 9 Nuclear Power Plant (NPP) designs. Each new NPP project benefits from previous projects and contains an increasing number of fully proven enhancements. In accordance with this evolutionary design approach, AECL has built on the Wolsong and Qinshan successes and the solid performance of the reference CANDU stations to define, review and implement the enhancements for the CANDU 9 NPP. Some of these enhancements include fully integrated project information systems and databases, safety enhancements coming from PSA studies and licensing activities, distributed control systems for plant-wide control and an advanced control center which addresses human factors engineering concepts. Examples of the Qinshan CANDU project delivery enhancements are the utilization of electronic engineering tools for the complete plant, and the linking of these tools with the project material management system and document management systems. The project information is reviewed and approved at the engineering office in Canada and then transmitted to site electronically. Once the electronic data is at site the information packages are extracted as necessary to enable construction and facilitate contract needs with minimum effort. This paper will provide details of the CANDU Qinshan project experiences as well as describing some of the corresponding CANDU 9 enhancements. (author)

Oxidation behavior analysis of cladding during severe accidents with combined codes for Qinshan Phase II Nuclear Power Plant

International Nuclear Information System (INIS)

Shi, Xingwei; Cao, Xinrong; Liu, Zhengzhi

2013-01-01

Highlights: • A new verified oxidation model of cladding has been added in Severe Accident Program (SAP). • A coupled analysis method utilizing RELAP5 and SAP codes has been developed and applied to analyze a SA caused by LBLOCA. • Analysis of cladding oxidation under a SA for Qinshan Phase II Nuclear Power Plant (QSP-II NPP) has been performed by SAP. • Estimation of the production of hydrogen has been achieved by coupled codes. - Abstract: Core behavior at a high temperature is extremely complicated during transition from Design Basic Accident (DBA) to the severe accident (SA) in Light Water Reactors (LWRs). The progression of core damage is strongly affected by the behavior of fuel cladding (oxidation, embrittlement and burst). A Severe Accident Program (SAP) is developed to simulate the process of fuel cladding oxidation, rupture and relocation of core debris based on the oxidation models of cladding, candling of melted material and mechanical slumping of core components. Relying on the thermal–hydraulic boundary parameters calculated by RELAP5 code, analysis of a SA caused by the large break loss-of-coolant accident (LBLOCA) without mitigating measures for Qinshan Phase II Nuclear Power Plant (QSP-II NPP) was performed by SAP for finding the key sequences of accidents, estimating the amount of hydrogen generation and oxidation behavior of the cladding

On application of the systematic approach to training in Qinshan NPP

International Nuclear Information System (INIS)

Wang Riqing

1997-01-01

The author describes the feature of systematic approach to training and introduces the situation about using the approach for training operation and maintenance personnel in Qinshan NPP. The final part of paper shows that there are still some problems worthy of serious consideration in application of the systematic approach to training in nuclear power plant

On the domestically-made heavy forging for reactor pressure vessels of PWR nuclear power plant

International Nuclear Information System (INIS)

Pan Xiren; Zhang Chen.

1988-01-01

The present situation of the foreign heavy forgings for nuclear reactor pressure vessels and the heavy forgings condition which is used for the Qinshan 300MWe nuclear power plant are described. Some opinions of domestic products is proposed

Bilateral cooperation and technology transfer between France and China at Daya-Bay, Qinshan II and Yibin

International Nuclear Information System (INIS)

Ma Fubanf; Zenf Wenxing; He Jiacheng; Charbonneau, S.; Darolles, J.F.; Ellia, G.; Freslon, H.

1994-01-01

The Daya-Bay nuclear power station in Guangdong Province, The Qinshan phase II nuclear power station in Zhejiang Province, and the fuel manufacturing facility at Yibin in Sichuan Province have all afforded Framatome the opportunity to develop wide-ranging bilateral cooperation and technology transfer with the People's Republic of china. These projects are all good examples of how a country with some nuclear power experience, such as the now-operating Qinshan 1 (300 M We) nuclear power unit designed and build by China itself, can make much more rapid progress in its civil nuclear power program through cooperation with an industry leader, such as Framatome

The database system of the real-time dose appreciation for Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Jiang Li; Chai Luquan

1993-01-01

The paper is about the data base system of the real-time dose appreciation for Qinshan Nuclear Power Plant and describes in detail the design of the system, the data structure, the programming and the characteristics

Application of project management in technology improvement of Qinshan III

International Nuclear Information System (INIS)

Liu Xiaonian

2008-01-01

During the operation of Qinshan III, many engineering modifications and renovation projects are being carried out. Advanced international project management methodologies accustomed to the policy and organizational characteristics of TQNPC were applied to the management of these projects. After practical application and development of these methodologies, the company finally sets up its own classification of project management system. The project management system is introduced and discussed for its evolving direction in this paper. (authors)

Tornado-resistance design for the nuclear safety structure of Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Xia Zufeng.

1987-01-01

The primary design consideration of anti-tornado of the nuclear safety structure of Qinshan Nuclear Power Plant is briefly presented. It mainly includes estimating the probability of tornado arising in the site, ascertaining the design requirments of the anti-tornado structures and deciding the tornado load acted on the structures

Preliminary site investigation for LL and IL radwaste disposal for Qinshan NPP

International Nuclear Information System (INIS)

Huang Yawen; Chen Zhangru

1993-01-01

With the purpose of selecting a disposal site for the low- and intermediate-level radwastes arising from Qinshan NPP, site investigations were carried out in several districts of Zhejiang Province. Investigation objectives included the circumstances of geology, hydrogeology, environmental ecology, and social economy. On the basis of collected data, five possible sites were recommended for policy-making reference and further investigation

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

CERN Document Server

Chiba, G

2002-01-01

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

Computerized nuclear material database management system for power reactors

International Nuclear Information System (INIS)

Cheng Binghao; Zhu Rongbao; Liu Daming; Cao Bin; Liu Ling; Tan Yajun; Jiang Jincai

1994-01-01

The software packages for nuclear material database management for power reactors are described. The database structure, data flow and model for management of the database are analysed. Also mentioned are the main functions and characterizations of the software packages, which are successfully installed and used at both the Daya Bay Nuclear Power Plant and the Qinshan Nuclear Power Plant for the purposed of handling nuclear material database automatically

The status of fast reactor technology development in China

International Nuclear Information System (INIS)

Xu Mi

1998-01-01

The paper will outline the main activities on fast reactor technology in China. In the year 1996, with the increasing of about 15 GWe installed electricity capacity, the total national electricity generation capacity has reached 225 GWe in the Country. Two nuclear power plants, Qinshan Phase 1 and Daya Bay have their rather good operation. The load factor of Qinshan Phase 1 was 84.7%. 76.1% and 64.1% for Daya Bay Unit 1 and Unit 2 respectively. During the Ninth 5-year (from 1996 to 2000) four NPPs Consisting of eight units of installed 6620MWe will be constructed. Under the framework of the High Technology Programme the Chinese Experimental Fast Reactor (CEFR) with the power 65MWth matched with 25MWe turbine-generator is still under preliminary design stage, which is sodium cooled pool type, (Pu,U)O 2 as fuel, in-core primary Went fuel storage, two mechanical pumps and four intermediate heat exchangers for primary circuit two loops for secondary circuits two independent immersed heat exchangers and air coolers with high stacks for passive residual heat removal system. Some design changes are presented in the paper. Concerning the R and D for the CEFR, besides the facilities already prepared, for demonstration of thermohydraulic characteristics of natural convection, a water simulation reactor pool facility in about one third scale is planned, in order to prepare the reactor physics experiments for its start-up, the zero power fast neutron facility with 50kg U-235 has been restored, for endurance testing of core subassemblies and getting some sodium loop operation experiences, Italian ESPRESSO and CEDI are under reconstruction in our lab. As for the engineering preparation of the project CEFR, the Feasibility Study Report was approved by Authorities on November last year. The site preparation and the design of incorporated to grid are just started. Finally, the activities of the international cooperation are presented in the paper. (author)

Qinshan plant display system: experience to date

International Nuclear Information System (INIS)

Bin, L.; Jiangdong, Y.; Weili, C.; Haidong, W.; Wangtian, L.; Lockwood, R.; Doucet, R.; Trask, D.; Judd, R.

2004-01-01

The two CANDU 6 units operated by the Third Qinshan Nuclear Power Corporation (TQNPC) include, as part of a control centre upgrade, a new plant display system (PDS). The PDS provides plant operators with new display and monitoring functionality designed to compliment the DCC capability. It includes new overview and trend displays (e.g., critical safety parameter monitor and user-defined trends), and enhanced annunciation based on AECL's Computerized Alarm Message List System (CAMLS) including an alarm interrogation capability. This paper presents a review of operating experience gained since the PDS was commissioned more than three years ago. It includes feedback provided by control room operators and trainers, PDS maintainers, and AECL development and support staff. It also includes an overview of improvements implemented since the PDS and suggestions for the future enhancements. (author)

A simulated test of physical starting and reactor physics on zero power facility of PWR

International Nuclear Information System (INIS)

Yao Zewu; Ji Huaxiang; Chen Zhicheng; Yao Zhiquan; Chen Chen; Li Yuwen

1995-01-01

The core neutron economics has been verified through experiments conducted at a zero power reactor with baffles of various thickness. A simulated test of physical starting of Qinshan PWR has been introduced. The feasibility and safety of the programme are verified. The research provides a valuable foundation for developing physical starting programme

Economic analysis of fuel management philosophy amendment in the second Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Cai Guangming

2006-01-01

In order to improve economic benefit, the Second Qinshan Nuclear Power Plant prepares to amend its fuel management philosophy after several fuel cycles. Economic evaluation is necessary before amendment of fuel management philosophy. Strong points and shortcomings are compared in this paper between yearly 1/4 refueling philosophy and 18 months refueling philosophy. (authors)

The reactor power control system based on digital control in nuclear power plant

International Nuclear Information System (INIS)

Liu Chong; Zhou Jianliang; Tan Ping

2010-01-01

The PLC (Programmable Logical Controller), digital communication and redundant techniques are applied in the rod control and position indication system(namely the reactor power control system) to perform the power control in the 300 MW reactor automatically and integrally in Qinshan Phase I project. This paper introduces the features, digital design methods of hardware of the instrumentation and control system (I and C) in the reactor power control. It is more convenient for the information exchange by human-machine interface (HMI), operation and maintenance, and the system reliability has been greatly improved after the project being reconstructed. (authors)

Study on the 90Sr absorption by agricultural plants grown in soil from Daya Bay and Qinshan area

International Nuclear Information System (INIS)

Zhao Wenhu; Hou Lanxin; Xu Shiming

1991-03-01

The soil around the Qinshan and Daya Bay nuclear power plants were used in this study. The 90 Sr was spread into the soil by irrigation. The amount of 90 Sr spread were 0.037, 0.37, 3.7 and 370 Bq per gram soil respectively. After being treated, the soil were employed to grow rice, wheat, rape, bean, asparagus lettuce, tomato and peas. The harvested plants were divided into seeds, stems and leaves, husks and roots to measure their radioactivity separately. The results showed that the amount of 90 Sr absorbed by the plants was directly proportional to the 90 Sr content in the soil. The absorbed 90 Sr was mainly distributed in the stems and leaves. The seeds absorbed the least amount of 90 Sr compared with the other portions. The old leaves absorbed greater 90 Sr than the buds. The accumulated 90 Sr per unit dry weight of all plants grown in the soil from Daya Bay area was greater than in the soil from Qinshan area. More than 80% of total 90 Sr was distributed in the top layer from 0 to 4 cm. The concentration factors of various plants were also given

The continual fuel management modification in Qinshan project II

International Nuclear Information System (INIS)

Ye Guodong; Pan Zefei; Zhang Xingtian

2010-01-01

The fuel management strategy is the basis of the nuclear power plants. The performance of the fuel management strategy affects the plants' safety and economy indicators directly. The paper summarizes all the modifications on the fuel management work in Qinshan Project II since the plant was established. It includes the surveillance system of physics tests, fetching in high performance fuel assemblies, reloading pattern optimization, and the modifications of the final safety analysis report. At the same time, it evaluates the benefit of the modifications in the few years. The experience in this paper is much helpful and could be implemented on the same type plants. (authors)

Radwastes management in Qinshan Nuclear power plants

International Nuclear Information System (INIS)

Zhou Huan; Ling Kechi; Wang Qingrong; Luo Jingfan

1987-01-01

The source terms input used as the basic data for designing the radwaste treatment systems of Qinshan Nuclear Power Plant [300 MW(e)] is presented. The classification of radioactive liquid wastes, off-gases and solid wastes, and their treatment techniques, as well as on-site storage facilities for solid wastes are described. For liquid waste, the method of filtration-evaporation-ion exchange will be used as the main treatment technique. For off-gas, Holdup-decay treatment will be used. For evaporator concentrates, indrumsolidification method with normal domestic portland cement will be used. The assessment of impact of effluents to environment at normal operation of the NPP is also made. The results show that it will be safe for inhabitants nearby during normal operation and it can meet the requirements of national standard ''Regulation of Radiation Protection''

Practice of radiation dose control for tech-modification items in Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Zhang Yong; Chen Zhongyu; Xu Hongming; Fan Liguang; Jiang Jianqi; Bu Weidong

2006-01-01

In order to improve the safety and reliability of nuclear power plant operation, many tech-modifications related to system or equipment have been completed since operation in Qinshan NPP. this paper introduces radiation dose control for mainly tech-modifications items related to radiation, including radiation protection optimization measures and experience in aspects of item planning, program writing, process control, etc. (authors)

Benchmark tests of JENDL-3.2 for thermal and fast reactors

International Nuclear Information System (INIS)

Takano, Hideki

1995-01-01

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

KIT/KPS of Qinshan phase-II and a discussion on integrated information management and automatic control

International Nuclear Information System (INIS)

Yan Changhui

2001-01-01

Centralized Data Processing and Safety Panel (KIT/KPS) of Qinshan Phase-II power project is described, and the necessity and engineering scheme is presented of integrated information management and automatic control that would achieve in power plant according to the technology scheme and technology trait of KIT/KPS

Manufacture of the 300 MW steam generator and pressure stabilizer for Qinshan Nuclear Power Station

International Nuclear Information System (INIS)

Qian Yi; Miao Deming.

1989-01-01

A brief description of the manufacturing process of the steam generator and pressure stabilizer for 300 MWe Qinshan Nuclear Power Station in Shanghai Boiler Works is presented, with special emphasis on fabrication facilities, test procedures and technological evaluations during the manufaturing process-imcluding deep driling of tubesheets, welding of tubes to tube-sheets and tube rolling tests

Brief introduction to project management of full scope simulator for Qinshan 300 MW Nuclear Power Unit

International Nuclear Information System (INIS)

Chen Jie

1996-01-01

The key points in development and engineering project management of full scope simulator for Qinshan 300 MW Nuclear Power Unit are briefly introduced. The Gantt chart, some project management methods and experience are presented. The key points analysis along with the project procedure will be useful to the similar project

Development of parallel 3D discrete ordinates transport program on JASMIN framework

International Nuclear Information System (INIS)

Cheng, T.; Wei, J.; Shen, H.; Zhong, B.; Deng, L.

2015-01-01

A parallel 3D discrete ordinates radiation transport code JSNT-S is developed, aiming at simulating real-world radiation shielding and reactor physics applications in a reasonable time. Through the patch-based domain partition algorithm, the memory requirement is shared among processors and a space-angle parallel sweeping algorithm is developed based on data-driven algorithm. Acceleration methods such as partial current rebalance are implemented. The correctness is proved through the VENUS-3 and other benchmark models. In the radiation shielding calculation of the Qinshan-II reactor pressure vessel model with 24.3 billion DoF, only 88 seconds is required and the overall parallel efficiency of 44% is achieved on 1536 CPU cores. (author)

Verification of results of core physics on-line simulation by NGFM code

International Nuclear Information System (INIS)

Zhao Yu; Cao Xinrong; Zhao Qiang

2008-01-01

Nodal Green's Function Method program NGFM/TNGFM has been trans- planted to windows system. The 2-D and 3-D benchmarks have been checked by this program. And the program has been used to check the results of QINSHAN-II reactor simulation. It is proved that the NGFM/TNGFM program is applicable for reactor core physics on-line simulation system. (authors)

The preparation and implementation of the commissioning of Qinshan nuclear power plant

International Nuclear Information System (INIS)

Huang Jinyuan

1993-05-01

The commissioning test of Qinshan Nuclear Power Plant is summarized. The preparation stage includes the organizations, commissioning programme, network planning, commissioning items, management procedures, responsibilities and interfaces between divisions, products ordering and supplying, personal training, quality assurance and the review and supervision by National Nuclear Safety Administration etc. The implementation stage includes the commissioning programme planning, intermediate hand-over inspection of the system and equipment, inspecting conditions and setting organizations for commissioning, the transition from commissioning to operating. Finally, some experiences in the commissioning test are presented in the article

The in-pile proving test for fuel assembly of Qinshan nuclear power plant

International Nuclear Information System (INIS)

Chen Dianshan; Zhang Shucheng; Kang Rixin; Wang Huarong; Chen Guanghan

1989-10-01

The in-pile proving test for fuel assembly of Qinshan nuclear power plant had been conducted in the experimental loop of HWRR at IAE (Institute of Atomic Energy) in Beijing, China, from January 1985 to December 1986. Average burnup of 27000 MWd/tU and peak burnup of 34000 MWd/tU of fuel rod had already been reached. The basic status of the experiment are described, emphasis is placed on the discussion of proving test parameters and analysis of experiment results

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

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

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

Optimization of reload core design for PWR and application to Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Shen Wei; Zhongsheng Xie; Banghua Yin

1995-01-01

A direct efficient optimization technique has been effected for automatically optimizing the reload of PWR. The objective functions include: maximization of end-of-cycle (EOC) reactivity and maximization of average discharge burnup. The fuel loading optimization and burnable poison (BP) optimization are separated into two stages by using Haling principle. In the first stage, the optimum fuel reloading pattern without BP is determined by the Linear Programming method using enrichments as control variable. In the second stage the optimum BP allocation is determined by the Flexible Tolerance Method using the number of BP rods as control variable. A practical and efficient PWR reloading optimization program based on above theory has been encoded and successfully applied to Qinshan Nuclear Power Plant(QNP)cycle 2 reloading design

JENDL-3.3 thermal reactor benchmark test

International Nuclear Information System (INIS)

Akie, Hiroshi

2001-01-01

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

Gray model prediction of the sea wall profile survey in the first process of Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Zang Deyan

1998-01-01

Based on gray system theory, the information about deformation observation of the first stage Qinshan nuclear power plant is analysed and predicted as well. The gray system theory is applied to engineering prediction and a large-scale building deformation observation. It is convenient to apply the model and it a has high degree of accuracy

Determination of total tritium in urine from residents living in the vicinity of nuclear power plants in Qinshan, China.

Science.gov (United States)

Shen, Bao-Ming; Ji, Yan-Qin; Tian, Qing; Shao, Xiang-Zhang; Yin, Liang-Liang; Su, Xu

2015-01-16

To estimate the tritium doses of the residents living in the vicinity of a nuclear power plant, urine samples of 34 adults were collected from residents living near the Qinshan nuclear power plant. The tritium-in-urine (HTO plus OBT) was measured by liquid scintillation counting. The doses of tritium-in-urine from participants living at 2, 10 and 22 km were in a range of 1.26-6.73 Bq/L, 1.31-3.09 Bq/L and 2.21-3.81 Bq/L, respectively, while the average activity concentrations of participants from the three groups were 3.53 ± 1.62, 2.09 ± 0.62 and 2.97 ± 0.78 Bq/L, respectively. The personal committed effective doses for males were 2.5 ± 1.7 nSv and for females they were 2.9 ± 1.3 nSv. These results indicate that tritium concentrations in urine samples from residents living at 2 km from a nuclear power plant are significantly higher than those at 10 km. It may be the downwind direction that caused a higher dose in participants living at 22 km. All the measured doses of tritium-in-urine are in a background level range.

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Manpower development for safe operation of nuclear power plant. China. Simulator software development. UNDP-Activity: 2.1.8-IAEA-Task-01. Technical report

International Nuclear Information System (INIS)

Feng, C.P.

1994-01-01

In the frameworks of the project ''manpower development for safe operation of nuclear power plant'' the development of reactor simulator software is described. Qinshan nuclear power plant was chosen as a reference one

Numerical forecast test on local wind fields at Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Chen Xiaoqiu

2005-01-01

Non-hydrostatic, full compressible atmospheric dynamics model is applied to perform numerical forecast test on local wind fields at Qinshan nuclear power plant, and prognostic data are compared with observed data for wind fields. The results show that the prognostic of wind speeds is better than that of wind directions as compared with observed results. As the whole, the results of prognostic wind field are consistent with meteorological observation data, 54% of wind speeds are within a factor of 1.5, about 61% of the deviation of wind direction within the 1.5 azimuth (≤33.75 degrees) in the first six hours. (authors)

The research on corrosion condition and anticorrosion methods of SEP system pipelines in Qinshan Nuclear Power Plant Phase II

International Nuclear Information System (INIS)

Zhang Wei; Cao Feng; Wang Jianjun

2010-01-01

SEP system in Qinshan nuclear power plant phase II provides drinking water and firefight water for nuclear island, conventional island, inner and outer of BOP structures. Many corrosion perforations in the SEP pipeline were found during operation. This article analysis the corrosion reasons and presents some reasonable treatment and surveillance methods. (authors)

Turbine and its turbine control system of full scope simulator for Qinshan 300 MW Nuclear Power Unit

International Nuclear Information System (INIS)

Zhang Dongwei; Zhu Jinping

1996-01-01

The simulation for Qinshan 300 MW Nuclear Power Unit turbine and turbine control system is briefly introduced. The simulation system includes lube oil system, jacking oil pump system, turning gear system, turbine supervisor system and turbine control system. It not only correctly simulates the process of turbine normal start up, operation, and shut down, but also the response of turbine under the malfunction conditions

Instructor station of full scope simulator for Qinshan 300 MW Nuclear Power Unit

International Nuclear Information System (INIS)

Wu Fanghui

1996-01-01

The instructor station of Full Scope Simulator for Qinshan 300 MW Nuclear Power Unit is based on SGI graphic workstation. The operation system is real time UNIX, and the development of man-machine interface, mainly depends on standard X window system, special for X TOOLKITS and MOTIF. The instructor station has been designed to increase training effectiveness and provide the most flexible environment possible to enhance its usefulness. Based on experiences in the development of the instructor station, many new features have been added including I/O panel diagrams, simulation diagrams, graphic operation of malfunction, remote function and I/O overrides etc

Multi-objective optimization of the reactor coolant system

International Nuclear Information System (INIS)

Chen Lei; Yan Changqi; Wang Jianjun

2014-01-01

Background: Weight and size are important criteria in evaluating the performance of a nuclear power plant. It is of great theoretical value and engineering significance to reduce the weight and volume of the components for a nuclear power plant by the optimization methodology. Purpose: In order to provide a new method for the optimization of nuclear power plant multi-objective, the concept of the non-dominated solution was introduced. Methods: Based on the parameters of Qinshan I nuclear power plant, the mathematical models of the reactor core, the reactor vessel, the main pipe, the pressurizer and the steam generator were built and verified. The sensitivity analyses were carried out to study the influences of the design variables on the objectives. A modified non-dominated sorting genetic algorithm was proposed and employed to optimize the weight and the volume of the reactor coolant system. Results: The results show that the component mathematical models are reliable, the modified non-dominated sorting generic algorithm is effective, and the reactor inlet temperature is the most important variable which influences the distribution of the non-dominated solutions. Conclusion: The optimization results could provide a reference to the design of such reactor coolant system. (authors)

China’s Nuclear Power Plants in Operation

Institute of Scientific and Technical Information of China (English)

无

2011-01-01

Qinshan Plant Phase I Located in Haiyan,Zhejiang Province,Qinshan Nuclear Power Plant Phase I is t he first 300-megawatt pressurized water reactor (PWR) nuclear power plant independently designed,constructed,operated and managed by China.The plant came into commercial operation in April 1994.

An investigation on technical bases of emergency plan zone determination of Qinshan Nuclear Power Base

International Nuclear Information System (INIS)

Duan Xuyi

2000-01-01

According to the general principal and the basic method of determination of emergency zone and safety criteria and in the light of the environmental and accidental release characteristic of Qinshan Nuclear Power Base, the expectation dose of assumed accident of each plant was compared and analyzed. In consideration of the impact factor of the size of emergency plan zone and referring to the information of emergency plan zone determination of other country in the world, the suggestions of determination method of emergency plan zone are proposed

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

International Nuclear Information System (INIS)

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

2012-01-01

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Qinshan phase II extension nuclear power project thermal stratification and fatigue stress analysis for pressurizer surge line

International Nuclear Information System (INIS)

Yu Xiaofei; Zhang Yixiong; Ai Honglei

2010-01-01

Thermal stratification of pressurizer surge line induced by the inside fluid brings on global bending moments, local thermal stresses, unexpected displacements and support loadings of the pipe system. In order to avoid a costly three-dimensional computation, a combined 1D/2D technique has been developed and implemented to analyze the thermal stratification and fatigue stress of pressurize surge line of QINSHAN Phase II Extension Nuclear Power Project in this paper, using the computer codes SYSTUS and ROCOCO. According to the mechanical analysis results of stratification, the maximum stress and cumulative usage factor, the loadings at connections of surge line to main pipe and RCP and the displacements of surge line at supports are obtained. (authors)

Major faults and troubleshooting for the power generator of Qinshan III

International Nuclear Information System (INIS)

Liu Guangming; Lu Yongfang; Wang Jun

2010-01-01

Generator faults can be sorted into 20 categories, mainly including water leakage, oil leakage, high temperature and short circuit, etc. The paper comprises two sections, the first section emphasizes on typical fault troubleshooting for power generator cooling water leakage, temperature rise and short circuit of Qinshan III, and the second section is conclusion. By expounding the troubleshooting for power generator cooling pipe leakage, -iron-core high temperature and rotor layer short circuit, the repair process and experience in the troubleshooting of typical fault including water leakage, temperature rise and short circuit are described in detail, so as to obtain the overall performance and parameters of the power generator, and provide useful means and plan for future troubleshooting. The paper can make reference to future troubleshooting for power generators. (authors)

Fast dose assessment models, parameters and code under accident conditions for Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Zhang, Z.Y.; Hu, E.B.; Meng, X.C.; Zhang, Y.; Yao, R.T.

1993-01-01

According to requirement of accident emergency plan for Qinshan Nuclear Power Plant, a Gaussian straight-line model was adopted for estimating radionuclide concentration in surface air. In addition, the effects of mountain body on atmospheric dispersion was considered. By combination of field atmospheric dispersion experiment and wind tunnel modeling test, necessary modifications have been done for some models and parameters. A computer code for assessment was written in Quick BASIC (V4.5) language. The radius of assessment region is 10 km and the code is applicable to early accident assessment. (1 tab.)

Optimization of the production plan and risk control in Third Qinshan Nuclear Power Co.,Ltd

International Nuclear Information System (INIS)

Zhou Jun

2009-01-01

Some optimized and improved measures have been taken in Third Qinshan Nuclear Power Co., Ltd. (TQNPC) to optimize the routine production plan management, strengthen the maintenance work risk analysis, and improve the plan execution capability. Which involve unified management of generation, refueling, periodic test and maintenance plans; simplifying the defect scale and reducing the intermediate link of defect treatment; intensifying the appraisal on plan execution and adopting star performance evaluation and merit rating measures. In this paper, the above-mentioned improvement and optimization are introduced comprehensively and systematically. (authors)

Analysis of Reasons for fluctuation in seal oil system on generator and countermeasures in Qinshan phase III project

International Nuclear Information System (INIS)

Jin Xiaodong

2012-01-01

Reasons for hydraulic differential fluctuations seal hydrogen oil on generator in Qinshan phase III project were analyzed, provide a basis for modifying Run method is to determine the causes and effects of seal oil flow changes and in the relationship between flow changes and hydraulic differential hydrogen oil changes according to reason Results were analyzed to adjust the running test, to verify the feasibility of running adjustment programs

Design and application of leakage monitor for reactor and control rod driving system

International Nuclear Information System (INIS)

Li, Dongyu; Zou, Yimin; Ling, Qiu; Guo, Lanying

2009-04-01

By measuring the number of γ photons produced by the annihilation of the β + particles of 13 N's decay product in the sample air, the nuclide density of 13 N can be obtained, comparing with its density in the reactor coolant, we can get the leakage information of the reactor vessel and control rod driving system, the article describes the cause of improvement in monitoring for leakage of reactor vessel and control rod driving system of Qinshan Second Nuclear Power Plant (PWR reactor), also the determination of monitoring method and system configuration, as well as the main technical index and function. Furthermore, the main parts and its function of the monitor are introduced. After operation for more than four years, it is proved that both the stability and MTBF index of the monitor meet the design, even more, thanks to the improvement of the algorithm, the Compton Effect caused by other nuclide became neglectable, the MDA of the monitor was lowered also. (authors)

The regeneration test of the secondary loop condensate polishing mixed bed resin in Qinshan NPP

International Nuclear Information System (INIS)

Xu Meijing; Dong Liming

1995-12-01

There are four condensate polishing mixed beds in the water chemical treatment plant of Qinshan NPP. 2125 kg of D001-TR type cation exchange resin, 2000 kg of D201-TR type anion exchange resin, and 375 kg of S-TR type inert resin are filled into each mixed bed. The bed height of resin is 1.2 m and the volume is about 2.7 m 3 . In order to regenerate the exhausted resin out of the bed, the pre-designed condensate polishing mixed bed regeneration process was used to regenerate the first exhausted resin. After the resin was scrubbed and separated, cation resin and anion resin were respectively regenerated, rinsed to resume the exchange capability of the resin. The regenerated mixed bed is able to keep higher efficiency for condensate polishing. The outlet water quality and the resin service-life are able to meet the design requirements or more favorable than that. During the test, some main cations and anions in the blow-off water at each procedure were analyzed. The analyzed results were used to make pre-designed regeneration process better. The test results proved that pre-designed process is reasonable and effective. (6 refs., 6 figs., 7 tabs.)

Architecture and design of third Qinshan nuclear power plant risk monitor

Energy Technology Data Exchange (ETDEWEB)

Wang, F.; Li, Y.; Wang, J.; Wang, J.; Hu, L. [Inst. of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, Univ. of Science and Technology of China, No.350 Shushanhu Road, Shushan District, Hefei, Anhui, 230031 (China)

2012-07-01

Risk monitor is a real-time analysis tool to determine the point-in-time risk based on actual plant configuration, which is an important application of PSA (Probabilistic Safety Assessment). In this study the status and development trend of risk monitor were investigated and a risk monitor named TQRM (Third Qinshan nuclear power plant Risk Monitor) was developed. The B/S architecture and the two key computing methods pre-solved and resolving PSA model method adopted in TQRM were introduced. The functions and technical features were also presented. Now TQRM has been on-line for more than one year and used in the operation and maintenance of TQNPP. The experience demonstrates that TQRM's results are accurate and real-time, the architecture is stable, and it could be extended and maintained conveniently for any other Risk-Informed Application. (authors)

A special device used for measuring waste gas flow rate in the vent channel of Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Zhang Yingjun; Zong Guifang; Shi Huaming; Yang Huimin; Jiang Yuana.

1988-01-01

A special Venturi-Pitot complex device is discribed which is used for measuring waste gas flow rate in the vent channel of Qinshan nuclear power plant. The device is located at the center of the channel. It can produce enlarged differential pressure signal under the condition of low gas velocity. And the flow resistance of this device is negligible. Experiments to determine the ratio of the velocity at the center of the channel to the average velocity were performed on a 1:12 test model. The special device was calibrated in a closed wind tunnel and its discharge coefficient was obtained. The uncertainty is ±3.5% and the nonlinearity is ±1.3%. The enlargement ratio and the discharge coefficient of the device are also deduced analytically on the basis of hydrodynamics theory

Macro testing for group constant library TPLIB-95

International Nuclear Information System (INIS)

Yao Dong; Zeng Daogui; Liu Jingbo; Wang Yingming; Li Huiyun

1996-04-01

A macro test of the group constant library TPLIB-95 was introduced. The TPLIB-95 is an updated group constant library created by China Nuclear Data Center for LWR fuel assembly calculation program package TPFAP based on the JENDL-3.1 evaluation nuclear data library. The calculations and analyses were carried out by using five thermal reactor benchmark issues, a set of PWR zero-power critical experiments, the first cycle reactor core of 300 MW Qinshan NPP as well as the first cycle reactor core of 900 MW Daya Bay NPP. The calculation results for the thermal reactor benchmark issues showed that the maximum deviation between the calculated and measured values for spectrum indexes is large, like 6.7% for ρ 28 of BAPL-2. However, the maximum deviation for k eff is only 0.29% for TRX-2. The calculation results for zero-power critical experiments showed that the calculated value of k eff obtained by using TPLIB-95 is closer to the measured value compared with the one obtained by using the original library TPLIB. The agreement between the calculated and measured values for critical boron concentration in the first cycle reactor cores in Qinshan NPP and Daya Bay NPP is quite good. The maximum deviation for the critical boron concentration is only 15 x 10 -6 /L. (8 figs., 5 tabs.)

Site ultrasonic measurement on RPV stud-bolt loading under hot transient of Qinshan NPP

International Nuclear Information System (INIS)

Qu Jiadi; Dou Yikang; Zhu Shiming

1994-08-01

It expounds that the key of solving thermal transient sealing problem is to obtain the thermal increment of stud-bolt loading. This loading, as a primary stress loading, is directly related to the bolt fatigue life and transient loading spectrum for vessel analysis. The fundamental works and main results of ultrasonic measurement on RPV stud-bolt loading on Qinshan site are also presented. The measuring capability has exceeded 1 m in length and temperature of 280 degree C, therefore, it is possible to be used in the field of NPP. The paper is the continuation of research work for sealing analysis and tests on the RPV (see SMiRT-9, 10)

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

International Nuclear Information System (INIS)

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

2003-01-01

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

TA-2 Water Boiler Reactor Decommissioning Project

International Nuclear Information System (INIS)

Durbin, M.E.; Montoya, G.M.

1991-06-01

This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m 3 of low-level solid radioactive waste and 35 m 3 of mixed waste. 15 refs., 25 figs., 3 tabs

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

International Nuclear Information System (INIS)

1966-01-01

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

137Cs absorption by growing rice planted in pot soil from Qinshan and Daya Bay area

International Nuclear Information System (INIS)

Shang Zhaorong; Yu Fengyi; Lu Zixian

1999-01-01

The pot experiment of growing rice contaminated with 137 Cs solution was designed as follows. (1) The same volume of 137 Cs solution was irrigated into rice soil from Guantang District around Qinshan NPP in seedling stage, booting stage and milk stage respectively with the same Specific Activity (SA) of 370 Bq/g soil , and the rice was sampled after maturity. (2) In the seedling stage, the rice cultured in the soil from Guantang District was irrigated by four different SA of 0.37, 3.7, 37 and 370 Bq/g soil respectively, and sampled after 30, 60 and 90 d. (3) Transfer Factors (TF) of edible parts of rice on five different soils were calculated for three different stage and four different 137 Cs levels. The results show that: 1) TF of Shenzhen soil is the highest with 1.86 in seed and 2.22 in stem and 4.05 in leaf, Changchuanba soil is the lowest with 0.09 in seed and 0.20 in stem and 0.20 in leaf, among the five different soils. 2) TF in milk stage is the highest with 0.46 in seed and 2.29 in stem and 2.87 in leaf, and booting stem is lowest with 0.09 in seed and 0.17 in stem and 0.17 in leaf, among the three different stage. 3) TF of soil with contamination in 0.37 Bq/g soil is the highest with 1.08 in seed and 3.70 in stem and 4.32 in leaf, and the contamination in 37 Bq/g soil is the lowest with 0.06 in seed and 0.10 in stem and 0.14 in leaf, among four different contamination levels

BR2 Reactor: Introduction

International Nuclear Information System (INIS)

Moons, F.

2007-01-01

The irradiations in the BR2 reactor are in collaboration with or at the request of third parties such as the European Commission, the IAEA, research centres and utilities, reactor vendors or fuel manufacturers. The reactor also contributes significantly to the production of radioisotopes for medical and industrial applications, to neutron silicon doping for the semiconductor industry and to scientific irradiations for universities. Along the ongoing programmes on fuel and materials development, several new irradiation devices are in use or in design. Amongst others a loop providing enhanced cooling for novel materials testing reactor fuel, a device for high temperature gas cooled fuel as well as a rig for the irradiation of metallurgical samples in a Pb-Bi environment. A full scale 3-D heterogeneous model of BR2 is available. The model describes the real hyperbolic arrangement of the reactor and includes the detailed 3-D space dependent distribution of the isotopic fuel depletion in the fuel elements. The model is validated on the reactivity measurements of several tens of BR2 operation cycles. The accurate calculations of the axial and radial distributions of the poisoning of the beryllium matrix by 3 He, 6 Li and 3T are verified on the measured reactivity losses used to predict the reactivity behavior for the coming decades. The model calculates the main functionals in reactor physics like: conventional thermal and equivalent fission neutron fluxes, number of displacements per atom, fission rate, thermal power characteristics as heat flux and linear power density, neutron/gamma heating, determination of the fission energy deposited in fuel plates/rods, neutron multiplication factor and fuel burn-up. For each reactor irradiation project, a detailed geometry model of the experimental device and of its neighborhood is developed. Neutron fluxes are predicted within approximately 10 percent in comparison with the dosimetry measurements. Fission rate, heat flux and

Performance test of condensate polishing system for Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

You Zhaojin; Qian Shijun; Lu Ruiting

1995-11-01

The flow chart, resin performance and water quality specifications of the condensate polishing system for Qinshan Nuclear Power Plant (QNPP) are briefly described. The initial regeneration process and the following service of the condensate polishing system are introduced. And the ability to remove corrosion products and ionic impurities of the condensate polishing system are verified during start-up, normal power operation and condenser leakage of the plant. The result shows that the performance of condensate polishing system in QNPP can completely meet the design requirements. Especially during the start-up of the unit or the leakage of the condenser, despite the inlet water quality of the polishers is far worse than the specified standard, the outlet water quality is still controlled within the indexes. Finally, several existing problems, such as 'volume ratio between resins is not optimum' and 'the inert resin and anion resin can not be stratified completely', in the condensate polishing system are also discussed. (4 refs., 1 fig., 8 tabs.)

The domestic development of rhodium self-powered detector used in the core of Qinshan third nuclear power plant

International Nuclear Information System (INIS)

Xiong Weihua; Zhang Zhenhua; Yu Yijun; Zhang Yun; Wu Jun; Deng Peng

2009-01-01

This article introduced Qinshan third nuclear power plant's Vanadium detector's principle of work, the domestically development's earlier period preparation, the craft processing process, the domestically sample's experiment as well as the sample in core demonstration test. Elaborated process of manufacture's quality control request and the essential craft, and the factory manufacture experiment situation, and to the installation and trial run process, the modification factor and the test result has carried on the introduction and the analysis. (authors)

EBR-2 [Experimental Breeder Reactor-2], IFR [Integral Fast Reactor] prototype testing programs

International Nuclear Information System (INIS)

Lehto, W.K.; Sackett, J.I.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.

1990-01-01

The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

Application of MCNPX 2.7.D for reactor core management at the research reactor BR2

International Nuclear Information System (INIS)

Kalcheva, Silva; Koonen, Edgar

2011-01-01

The paper discusses application of the Monte Carlo burn up code MCNPX 2.7.D for whole core criticality and depletion analysis of the Material Testing Research Reactor BR2 at SCK-CEN in Mol, Belgium. Two different approaches in the use of MCNPX 2.7.D are presented. The first methodology couples the evolution of fuel depletion, evaluated by MCNPX 2.7.D in an infinite lattice with a steady-state 3-D power distribution in the full core model. The second method represents fully automatic whole core depletion and criticality calculations in the detailed 3-D heterogeneous geometry model of the BR2 reactor. The accuracy of the method and computational time as function of the number of used unique burn up materials in the model are being studied. The depletion capabilities of MCNPX 2.7.D are compared vs. the developed at the BR2 reactor department MCNPX & ORIGEN-S combined method. Testing of MCNPX 2.7.D on the criticality measurements at the BR2 reactor is presented. (author)

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

International Nuclear Information System (INIS)

2015-01-01

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

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

International Nuclear Information System (INIS)

Takano, Hideki; Kaneko, Kunio.

1989-10-01

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

Plasma-catalyst hybrid reactor with CeO2/γ-Al2O3 for benzene decomposition with synergetic effect and nano particle by-product reduction.

Science.gov (United States)

Mao, Lingai; Chen, Zhizong; Wu, Xinyue; Tang, Xiujuan; Yao, Shuiliang; Zhang, Xuming; Jiang, Boqiong; Han, Jingyi; Wu, Zuliang; Lu, Hao; Nozaki, Tomohiro

2018-04-05

A dielectric barrier discharge (DBD) catalyst hybrid reactor with CeO 2 /γ-Al 2 O 3 catalyst balls was investigated for benzene decomposition at atmospheric pressure and 30 °C. At an energy density of 37-40 J/L, benzene decomposition was as high as 92.5% when using the hybrid reactor with 5.0wt%CeO 2 /γ-Al 2 O 3 ; while it was 10%-20% when using a normal DBD reactor without a catalyst. Benzene decomposition using the hybrid reactor was almost the same as that using an O 3 catalyst reactor with the same CeO 2 /γ-Al 2 O 3 catalyst, indicating that O 3 plays a key role in the benzene decomposition. Fourier transform infrared spectroscopy analysis showed that O 3 adsorption on CeO 2 /γ-Al 2 O 3 promotes the production of adsorbed O 2 - and O 2 2‒ , which contribute benzene decomposition over heterogeneous catalysts. Nano particles as by-products (phenol and 1,4-benzoquinone) from benzene decomposition can be significantly reduced using the CeO 2 /γ-Al 2 O 3 catalyst. H 2 O inhibits benzene decomposition; however, it improves CO 2 selectivity. The deactivated CeO 2 /γ-Al 2 O 3 catalyst can be regenerated by performing discharges at 100 °C and 192-204 J/L. The decomposition mechanism of benzene over CeO 2 /γ-Al 2 O 3 catalyst was proposed. Copyright © 2017 Elsevier B.V. All rights reserved.

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

International Nuclear Information System (INIS)

Talamo, Alberto; Gudowski, Waclaw

2005-01-01

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

Simulation of control performance under house load transients for nuclear power plant

International Nuclear Information System (INIS)

Liao Zhongyue; Wang Yuanlong; Tang Yuyuan; Liu Jiong

1999-01-01

The CATIA2 code is used to simulate the extreme normal transients--house load transients of Qinshan Phase II 600 MW nuclear power plant. The simulating results show that all of the reactor main parameters are operating in the allowable ranges, the reactor system is stable, and the control characteristics of the nuclear power plant is satisfactory. They are also good in agreement with Framatome's results

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)

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

International Nuclear Information System (INIS)

Liu Wenjin; Zeng Zhongxiu; Ye Xianhui; Wu Wanjun

2013-01-01

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

Efficient H2O2/CH3COOH oxidative desulfurization/denitrification of liquid fuels in sonochemical flow-reactors.

Science.gov (United States)

Calcio Gaudino, Emanuela; Carnaroglio, Diego; Boffa, Luisa; Cravotto, Giancarlo; Moreira, Elizabeth M; Nunes, Matheus A G; Dressler, Valderi L; Flores, Erico M M

2014-01-01

The oxidative desulfurization/denitrification of liquid fuels has been widely investigated as an alternative or complement to common catalytic hydrorefining. In this process, all oxidation reactions occur in the heterogeneous phase (the oil and the polar phase containing the oxidant) and therefore the optimization of mass and heat transfer is of crucial importance to enhancing the oxidation rate. This goal can be achieved by performing the reaction in suitable ultrasound (US) reactors. In fact, flow and loop US reactors stand out above classic batch US reactors thanks to their greater efficiency and flexibility as well as lower energy consumption. This paper describes an efficient sonochemical oxidation with H2O2/CH3COOH at flow rates ranging from 60 to 800 ml/min of both a model compound, dibenzotiophene (DBT), and of a mild hydro-treated diesel feedstock. Four different commercially available US loop reactors (single and multi-probe) were tested, two of which were developed in the authors' laboratory. Full DBT oxidation and efficient diesel feedstock desulfurization/denitrification were observed after the separation of the polar oxidized S/N-containing compounds (S≤5 ppmw, N≤1 ppmw). Our studies confirm that high-throughput US applications benefit greatly from flow-reactors. Copyright © 2013 Elsevier B.V. All rights reserved.

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

Directory of Open Access Journals (Sweden)

New Pei Yee

2008-04-01

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

Reliability improvement of potential transformer and secondary circuit of 6 kV 1E-class buses in Qinshan nuclear power plant

International Nuclear Information System (INIS)

Qian Houjun

2014-01-01

There are design defects in potential transformer (PT) and secondary circuit of 6 kV 1E-class buses in Qinshan Nuclear Power Plant Nuclear Island. During the operating period, there happened several serious operational events (loss of power) caused by PT resonance. The essay analyses the defects of original design, and put forward corresponding modification measures, which have been carried out by two steps between 2009 and 2010, and after the modification the same problems have not happened again. (author)

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1966-07-01

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

Core monitoring at the WNP-2 reactor

International Nuclear Information System (INIS)

Skeen, D.R.; Torres, R.H.; Burke, W.J.; Jenkins, I.; Jones, S.W.

1992-01-01

The WNP-2 reactor is a 3,323-MW(thermal) boiling water reactor (BWR) that is operated by the Washington Public Power Supply System. The WNP-2 reactor began commercial operation in 1984 and is currently in its eighth cycle. The core monitoring system used for the first cycle of operation was supplied by the reactor vendor. Cycles 2 through 6 were monitored with the POWERPLEX Core Monitoring Software System (CMSS) using the XTGBWR simulation code. In 1991, the supply system upgraded the core monitoring system by installing the POWERPLEX 2 CMSS prior to the seventh cycle of operation for WNP-2. The POWERPLEX 2 CMSS was developed by Siemens Power Corporation (SPC) and is based on SPC's advanced state-of-the-art reactor simulator code MICROBURN-B. The improvements in the POWERPLEX 2 system are possible as a result of advances in minicomputer hardware

EBR-2 [Experimental Breeder Reactor-2] test programs

International Nuclear Information System (INIS)

Sackett, J.I.; Lehto, W.K.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.; Hill, D.J.

1990-01-01

The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development, (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development, advanced control system development, plant diagnostics development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

Dynamics of TRIGA-3 Salazar Reactor

International Nuclear Information System (INIS)

Gallardo S, L.F.

1990-01-01

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

Reactor BR2

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2000-07-01

The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported.

Reactor BR2

International Nuclear Information System (INIS)

Gubel, P.

2000-01-01

The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported

Processing requirements for property optimization of Eu2O3-W cermets for fast reactor neutron absorber applications

International Nuclear Information System (INIS)

Pasto, A.E.; Tennery, V.J.

1977-01-01

Europium sesquioxide is a candidate fast reactor neutron absorber material. It possesses several desirable characteristics for this application, but has a low thermal conductivity. This gives rise to pellet cracking during reactor operation. To increase the thermal conductivity without great sacrifice in nuclear worth, addition of tungsten to Eu 2 O 3 has been evaluated. Synthesis and fabrication techniques described allow preparation of high density compacts of Eu 2 O 3 -15 vol. percent tungsten, possessing favorable thermal conductivity and thermal expansion characteristics

Presence of Tritium in the Cooling Circuits of the Reactors G2 and G3; Presence de tritium dans les circuits de refroidissement des reacteurs G2 et G3

Energy Technology Data Exchange (ETDEWEB)

Estournel, R [Commissariat a l' Energie Atomique. Centre de Production de Plutonium de Marcoule, 30 - Chusclan (France)

1962-07-01

In a reactor of the G 2-G 3 type, tritium can be formed by the neutronic bombardment of many elements present in the core. Tritium was found to be present in the cooling circuits of the reactors G 2 and G 3 in the water coming from the regeneration of the CO{sub 2} dehydrating columns. (author) [French] Dans un reacteur du type G 2 - G 3, le tritium peut etre forme par le bombardement. neutronique de nombreux elements existant dans le c r. La presence de tritium dans les circuits de refroidissement des reacteurs G 2 - G 3 a ete mis en evidence dans l'eau provenant de la regeneration des colonnes de deshydratation du CO{sub 2}. (auteur)

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

International Nuclear Information System (INIS)

Kishimoto, Katsumi; Arigane, Kenji

2005-03-01

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

Active species in a large volume N2-O2 post-discharge reactor

International Nuclear Information System (INIS)

Kutasi, K; Pintassilgo, C D; Loureiro, J; Coelho, P J

2007-01-01

A large volume post-discharge reactor placed downstream from a flowing N 2 -O 2 microwave discharge is modelled using a three-dimensional hydrodynamic model. The density distributions of the most populated active species present in the reactor-O( 3 P), O 2 (a 1 Δ g ), O 2 (b 1 Σ g + ), NO(X 2 Π), NO(A 2 Σ + ), NO(B 2 Π), NO 2 (X), O 3 , O 2 (X 3 Σ g - ) and N( 4 S)-are calculated and the main source and loss processes for each species are identified for two discharge conditions: (i) p = 2 Torr, f = 2450 MHz, and (ii) p = 8 Torr, f = 915 MHz; in the case of a N 2 -2%O 2 mixture composition and gas flow rate of 2 x 10 3 sccm. The modification of the species relative densities by changing the oxygen percentage in the initial gas mixture composition, in the 0.2%-5% range, are presented. The possible tuning of the species concentrations in the reactor by changing the size of the connecting afterglow tube between the active discharge and the large post-discharge reactor is investigated as well

MCNP apply in calculating reactor critical coefficient Keff under the changing of the burnable poison rod

International Nuclear Information System (INIS)

Wang Xinghua; Zhou Sichun; Zhang Qingxian; Zhao Feng; Liu Jun; Zhu Jian

2013-01-01

Taking Qinshan nuclear power plant as an example, in this paper, Monte Carlo method was used in the MCNP procedures for the establishment of nuclear power station simulation model, construct the reactor pressure vessel and vessel core component composition and arrangement, KCODE card was used to calculate the effect of the number and the location of burnable poison control rod factor K eff by the boron acid. The calculation results show that, with the increasing in the number of burnable poison control rod value-added factor K eff shown a downward trend, and with the burnable poison control rod from the dense to sparse, which K eff will be decreasing slowly. This condition is consistent with the theoretical. (authors)

CANDU flexible and economical fuel technology in China

Energy Technology Data Exchange (ETDEWEB)

Mingjun, C. [CNNC Nuclear Power Operation Management Co., Zhejiang (China); Zhenhua, Z.; Zhiliang, M. [CNNC Third Qinshan Nuclear Power Co., Zhejiang (China); Cottrell, C.M.; Kuran, S. [Candu Energy Inc., Mississauga, ON (Canada)

2014-07-01

Use in CANDU reactor is one good option of recycled uranium (RU) and thorium (Th) resource. It is also good economy to CANDU fuel. Since 2008 Qinshan CANDU Plant and our partners (Candu Energy and CNNC and NPIC) have made great efforts to develop the engineering technologies of Flexible and Economical Fuel (RU and Th) in CANDU type reactor and finding the CANDU's position in Chinese closed fuel cycle (CFC) system. This paper presents a proposal of developing strategy and implementation plan. Qinshan CANDU reactors will be converted to use recycled and depleted uranium based fuels, a first-of-its-kind. The fuel is composed of both recycled and depleted uranium and simulating natural uranium behavior. This paper discusses its development, design, manufacture and verification tested with success and the full core implementation plan by the end of 2014. (author)

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

Reactor building pressure proof test (PPT) and leak rate test (LRT) of Qinshan phase III (CANDU) project

International Nuclear Information System (INIS)

Gu Jun; Shi Jinqi; Fan Fuping

2004-12-01

As the first reactor building (R/B) without stainless steel liner in china, TQNPC studied the containment characteristics, such as strong concrete absorb/release air effect, poor containment penetration. etc. And carefully prepared test scheme and emergency response, creatively introduced the instrument air self-supply system in reactor building, developed the special measurement and analysis system for PPT and LRT, organized work under high-pressure on large-scale in the test. Finally got the containment leak rate result and the test-cost-time value is the best in all same type tests. (authors)

Neutron flux determinations in the reactors G2 and G3 during operation; Releves du flux neutronique dans les reacteurs G2 et G3 en puissance

Energy Technology Data Exchange (ETDEWEB)

Boulinier, C; Faurot, P; Sagot, M; Teste du Bailler, A [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1961-07-01

After demonstrating the sensitivity of the distribution of power in a production reactor to a deformation caused by dissymmetries of reactivity in the reactor, the authors describe the method of neutron flux determination devised for the reactors G2 and G3 under working conditions; the detector used is a tungsten or nickel wire, the {gamma} activity of which is measured with an ionisation chamber. Several flux determinations are given as examples to illustrate the sensitivity of the method. (author) [French] Apres avoir mis en evidence la sensibilite de la repartition de la puissance dans un reacteur de production a une deformation provoquee par de faibles dissymetries de reactivite dans le reacteur, les auteurs decrivent la methode de releve du flux neutronique mise au point pour les reacteurs G2 et G3 en puissance; le detecteur utilise est un fil de tungstene ou de nickel dont l'activite {gamma} est mesuree a l'aide d'une chambre d'ionisation. Quelques releves de flux illustrant la sensibilite de la methode sont donnes a titre d'exemple. (auteur)

The dynamic response of the containment of the Qinshan nuclear power plant to the aircraft impact loading

International Nuclear Information System (INIS)

Zuo Jiahong; Han Liangbi; Xia Zufeng

1991-08-01

The structural response of the containment of the Qinshan Nuclear Power Plant under the standard-load-function to aircraft impact has been analyzed by using the ADINA code considering an axisymmetric continuum model, which is assumed a mixed-model for the steel-concrete mixture. It consists of 179 four-node isoparametric concrete elements and 118 steel elements. In order to obtain optimum results, the nonlinear behavior of materials and structures, dynamic modes of failure and damage have been considered in the numerical solution. The coordinate system is based on the total Lagrangian formulation. The F.E. system has been solved using an incremental interactions (BFGS method) with 600 steps totally. A discussion of the overall behavior of the containment for the aircraft impact loading, especially the nonlinear behavior of the local impacted area is presented

Reactor safety issues resolved by the 2D/3D Program

International Nuclear Information System (INIS)

Damerell, P.S.; Simons, J.W.

1993-07-01

The 2D/3D Program studied multidimensional thermal-hydraulics in a PWR core and primary system during the end-of-blowdown and post-blowdown phases of a large-break LOCA (LBLOCA), and during selected small-break LOCA (SBLOCA) transients. The program included tests at the Cylindrical Core Test Facility (CCTF), the Slab Core Test Facility (SCTF), and the Upper Plenum Test Facility (UPTF), and computer analyses using TRAC. Tests at CCTF investigated core thermal-hydraulics and overall system behavior while tests at SCTF concentrated on multidimensional core thermal-hydraulics. The UPTF tests investigated two-phase flow behavior in the downcomer, upper plenum, tie plate region, and primary loops. TRAC analyses evaluated thermal-hydraulic behavior throughout the primary system in tests as well as in PWRs. This report summarizes the test and analysis results in each of the main areas where improved information was obtained in the 2D/3D Program. The discussion is organized in terms of the reactor safety issues investigated

The CO{sub 2} cooling gas for the reactors G2/G3 (leaking, analysis, activity); Le CO{sub 2} de refroidissement des reacteurs G2/G3 (fuites, analyse, activite)

Energy Technology Data Exchange (ETDEWEB)

Meiffren, J; Dupay, F [Commissariat a l' Energie Atomique, Centre de Production de Plutonium, Marcoule (France). Centre d' Etudes Nucleaires

1965-07-01

The main objective of this study is to publicise the data obtained during five years operation of the reactor G2 and G3 at Marcoule as far as the cooling gas is concerned, from storage of reserves up to its slow escape into the atmosphere, and including all the stages of its practical use, its chemical examination, its nuclear behaviour and its possible physicochemical transformation. This work can not only yield information about the operations carried out at Marcoule but can also provide useful suggestions for improving the sealing and for decreasing the activity of the pressurized gas circuits in reactors similar to G2/G3. (authors) [French] Le but principal de cette etude est de diffuser les connaissances acquises au cours de cinq annees d'exploitation des reacteurs G2 et G3 de Marcoule en ce qui concerne le gaz de refroidissement, depuis son stockage d'appoint jusqu'a son echappement lent dans l'atmosphere, en passant par tous les stades de son utilisation pratique, de son etude chimique, de son comportement nucleaire, eventuellement de ses transformations physico-chimiques. Cette etude peut, non seulement renseigner sur les operations effectuees couramment a Marcoule, mais egalement donner des suggestions interessantes pour l'amelioration de l'etancheite et la diminution de l'activite des circuits de gaz en pression dans des reacteurs analogues a G2/G3. (auteurs)

Radiation protection at the RA Reactor in 1988, Part -2, RA reactor annual report

International Nuclear Information System (INIS)

Ninkovic, M.; Ajdacic, N.; Zaric, M.; Vukovic, Z.

1988-01-01

Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor and radiation protection; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Decontamination and relevant actions, collecting and treatment of fluid effluents; and and solid radioactive wastes [sr

Hydrogenation of furfural at the dynamic Cu surface of CuOCeO2/Al2O3 in vapor phase packed bed reactor

Science.gov (United States)

The hydrogenation of furfural to furfuryl alcohol over a CuOCeO2/'-Al2O3 catalyst in a flow reactor is reported. The catalyst was prepared by the wet impregnation of Cu onto a CeO2/'-Al2O3 precursor. The calcined catalyst was then treated with HNO3 to remove surface CuO resulting in a mixed CuCe oxi...

Neutron irradiation damage in Al2O3 and Y2O3

International Nuclear Information System (INIS)

Clinard, F.W. Jr.; Bunch, J.M.; Ranken, W.A.

1975-01-01

Two ceramics under consideration for use in fusion reactors, Al 2 O 3 and Y 2 O 3 , were irradiated in the EBR-II fission reactor at 650, 875, and 1025 0 K to fluences between 2 and 6 x 10 21 n/cm 2 (E greater than 0.1 MeV). Samples evaluated include sapphire, Lucalox, alumina, Y 2 O 3 , and Y 2 O 3 -10 percent ZrO 2 (Yttralox). All Al 2 O 3 specimens swelled significantly (1 to 3 percent), with most of the growth observed in sapphire along the c-axis at the higher temperatures. Al 2 O 3 samples irradiated at 875 to 1025 0 K contained a high density of small aligned ''pores''. Irradiated Y 2 O 3 -based ceramics exhibited dimensional stability and a defect content consisting primarily of unresolved damage and/or dislocation loops. The behavior of these ceramics under irradiation is discussed, and the relevance of fission neutron damage studies to fusion reactor applications is considered. (auth)

A packed bed membrane reactor for the oxidative dehydrogenation of propane on a Ga2O3 / MoO3 based catalyst

NARCIS (Netherlands)

Kotanjac, Ž.S.; Sint Annaland, van M.; Kuipers, J.A.M.

2010-01-01

Oxidative dehydrogenation of propane has been studied over a Ga2O3/MoO3 based catalyst. Using a differentially operated packed bed reactor with premixed oxygen and propane feed, the kinetic parameters for the main reaction and the consecutive and parallel reactions were experimentally determined. It

Integral test of JENDL-3.3 for thermal reactors

International Nuclear Information System (INIS)

Okumura, Keisuke; Mori, Takamasa

2003-01-01

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

Radiation protection at the RA Reactor in 1998, RA reactor annual report, Part -2

International Nuclear Information System (INIS)

Ninkovic, M.; Pavlovic, R.; Mandic, M.; Pavlovic, S.; Grsic, Z.

1998-01-01

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

EL3 reactor description and safety analysis report

International Nuclear Information System (INIS)

1969-02-01

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

Decommissioning of a small reactor (BR3 reactor, Belgium)

International Nuclear Information System (INIS)

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

2002-01-01

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

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

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

International Nuclear Information System (INIS)

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

1992-09-01

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

Optimized Control Rods of the BR2 Reactor

Energy Technology Data Exchange (ETDEWEB)

Kalcheva, Silva; Koonen, E.

2007-09-15

At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

Optimized Control Rods of the BR2 Reactor

International Nuclear Information System (INIS)

Kalcheva, Silva; Koonen, E.

2007-01-01

At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

Conclusions of the experts group of the RA reactor at the meeting held on November 2 and 3 1964 - Annex 12a

International Nuclear Information System (INIS)

Pavicevic, M.

1964-01-01

Conclusions of the experts group of the RA reactor are related to: analyses of reactor operation at 6.5 MW power with heavy water coolant flow of 250 m 3 /h (2 pumps rotation speed 1500 rotations/min); decisions of future operation; further preparation activities related to reactor operation in forced regime and reduced cooling conditions

EL-3 dismantling of an experimental reactor

International Nuclear Information System (INIS)

1989-01-01

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

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)

Study on partial overheat of the isolated phase busbar outlet box in Qinshan NPP phase Ⅱ

International Nuclear Information System (INIS)

Tang Fangxuan; Zhang Jian; Zeng Limin; Bao Yanxing; Zhang Lie; Yang Yuemin

2013-01-01

This paper recommended the structure of the isolated phase busbar outlet box installed in Qinshan II. The study on partial overheat of the outlet box shows that the ultimate causes are the loss of concentrated eddy current and short of cooling. So the improvement principles of 'distributing eddy current, cutting off inductive circle current and strengthening of ventilation' were determined. A new structure test outlet box was designed and manufactured, and the temperature rising experiment was carried out. Some alterations were made in the new structure outlet box, e.g. isolating materials were added between side plates of the upper outlet box, and also between the upper and lower outlet box. Two cooling blowers were added to the upper outlet box. After putting into operation, the hot-spot temperature of the new outlet box was greatly lowered down. Thus the operation environment was improved, and the operation safety ensured. It can be useful references for analyzing and dealing with similar problems. (authors)

Variation of radioactivity in the environmental media and dose evaluation in Suzhou city after normal operation of Qinshan Nuclear Power Station condition

International Nuclear Information System (INIS)

Fu Rongchu; Liu Li

2002-01-01

Objective: To study the radioactive monitoring in environmental media of Suzhou City when Qinshan Nuclear Power Station was in normal operational condition (from 1992-2001). Methods: The radiochemical method was used for monitoring the radioactivity level in air, soil and food. Results: The total radioactivity, concentrations of 134 I and 134,137 Cs in environmental media was far lower than the limit values specified by the national standard GB. Conclusion: The radioactivity level in Suzhou City is at the natural background level. The individual annual average effective dose for adults in that period caused by ingestion 134,137 Cs in food is 4.41 x 10 -4 mSv/a

Comparison and analysis of 1D/2D/3D neutronics modeling for a fusion reactor

International Nuclear Information System (INIS)

Li, J.; Zeng, Q.; Chen, M.; Jiang, J.; Wu, Y.

2007-01-01

During the course of analyzing the characteristics for fusion reactors, the refined calculations are needed to confirm that the nuclear design requirements are met. Since the long computational time is consumed, the refined three-dimensional (3D) representation has been used primarily for establishing the baseline reference values, analyzing problems which cannot be reduced by symmetry considerations to lower dimensions, or where a high level of accuracy is desired locally. The two-dimensional (2D) or one-dimensional (1D) description leads itself readily to resolve many problems, such as the studies for the material fraction optimization, or for the blanket size optimization. The purpose of this paper is to find out the differences among different geometric descriptions, which can guide the way to approximate and simplify the computational model. The fusion power reactor named FDS-II was designed as an advanced fusion power reactor to demonstrate and validate the commercialization of fusion power by Institute of Plasma Physics, Chinese Academy of Science. In this contribution, the dual-cooled lithium lead (DLL) blanket of FDS-II was used as a reference for neutronics comparisons and analyses. The geometric descriptions include 1D concentric sphere model, 1D, 2D and 3D cylinder models. The home-developed multi-functional neutronics analysis code system VisualBUS, the Monte Carlo transport code MCNP and nuclear data library HENDL have been used for these analyses. The neutron wall loading distribution, tritium breeding ratio (TBR) and nuclear heat were calculated to evaluate the nuclear performance. The 3D calculation has been used as a comparison reference because it has the least errors in the treatment of geometry. It is suggested that the value of TBR calculated by the 1D approach should be greater than 1.3 to satisfy the practical need of tritium self-sufficiency. The distribution of nuclear heat based on the 2D and 3D models were similar since they all consider

PARR-2: reactor description and experiments

International Nuclear Information System (INIS)

Wyne, M.F.; Meghji, J.H.

1990-12-01

PARR-2 is a miniature neutron source reactor (MNSR) research reactor has been designed at the rate of 27 kW. Reactor assembly comprises of peaking characteristics with a self limiting flux. In this report reactor description with its assembly and instrumentation control system has been explained. The reactor engineering and physics experiments which can be performed on this reactor are explained in this report. PARR-2 is fueled with HEU fuel pins which are about 90% enriched in U-235. Specific requirements for the safety of the reactor, its building and the personnel, normal instrumentation as required in an industrial environment is sufficient. (A.B.)

Reactor BR2: Introduction

International Nuclear Information System (INIS)

Gubel, P.

2000-01-01

The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. A safety audit was conduced by the IAEA, the conclusions of which demonstrated the excellent performance of the plant in terms of operational safety. In 1999, the CALLISTO facility was extensively used for various programmes involving LWR pressure vessel materials, IASCC of LWR structural materials, fusion reactor materials and martensic steels for use in ADS systems. In 1999, BR2's commercial programmes were further developed

Material test reactor fuel research at the BR2 reactor

Energy Technology Data Exchange (ETDEWEB)

Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

2012-03-15

The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

A New Application of Support Vector Machine Method: Condition Monitoring and Analysis of Reactor Coolant Pump

International Nuclear Information System (INIS)

Meng Qinghu; Meng Qingfeng; Feng Wuwei

2012-01-01

Fukushima nuclear power plant accident caused huge losses and pollution and it showed that the reactor coolant pump is very important in a nuclear power plant. Therefore, to keep the safety and reliability, the condition of the coolant pump needs to be online condition monitored and fault analyzed. In this paper, condition monitoring and analysis based on support vector machine (SVM) is proposed. This method is just to aim at the small sample studies such as reactor coolant pump. Both experiment data and field data are analyzed. In order to eliminate the noise and useless frequency, these data are disposed through a multi-band FIR filter. After that, a fault feature selection method based on principal component analysis is proposed. The related variable quantity is changed into unrelated variable quantity, and the dimension is descended. Then the SVM method is used to separate different fault characteristics. Firstly, this method is used as a two-kind classifier to separate each two different running conditions. Then the SVM is used as a multiple classifier to separate all of the different condition types. The SVM could separate these conditions successfully. After that, software based on SVM was designed for reactor coolant pump condition analysis. This software is installed on the reactor plant control system of Qinshan nuclear power plant in China. It could monitor the online data and find the pump mechanical fault automatically.

In-core fuel management activities in China

International Nuclear Information System (INIS)

Ruan Keqiang; Chen Renji; Hu Chuanwen

1990-01-01

The development of nuclear power in China has reached such a stage that PWR in-core fuel management becomes an urgent problem. At present the main effort is concentrated on solving the Qinshan nuclear power plant and Daya Bay nuclear power plant fuel management problems. For the Qinshan PWR (300 MWe) two packages of in-core fuel management code were developed, one with simplified nodal diffusion method and the other uses advanced Green's function nodal method. Both were used in the PWR core design. With the help of the two code packages first two cycles of the Qinshan PWR core burn-up were calculated. Besides, several research works are under way in the following areas: improvement of the nodal diffusion method and other coarse mesh method in terms of computing speed and accuracy; backward diffusion technique for fuel management application; optimization technique in the fuel loading pattern searching. As for the Daya Bay PWR plant (twin 900 MWe unit), the problem about using what kind of code package for in-core fuel management is still under discussion. In principle the above mentioned code packages are also applicable to it. Besides PWR, in-core fuel management research works are also under way for research reactors, for example, heavy water research reactor and high flux research reactor in some institutes in China. China also takes active participation in international in-core fuel management activities. (author). 19 refs

Reactor BR2. Introduction

International Nuclear Information System (INIS)

Gubel, P.

2001-01-01

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

Reactor BR2. Introduction

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2001-04-01

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

Integral test of JENDL-3.3 for fast reactors

International Nuclear Information System (INIS)

Chiba, Gou

2003-01-01

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

Mineralization and defluoridation of 2,2,3,3-tetrafluoro -1-propanol (TFP) by UV oxidation in a novel three-phase fluidized bed reactor (3P-FBR).

Science.gov (United States)

Shih, Yu-Jen; Tsai, Meng-Tso; Huang, Yao-Hui

2013-05-01

2,2,3,3-Tetrafluoro-1-propanol (TFP, C3H4F4O, M.W. = 132.06) is extensively used as the solvent in CD-R and DVD-R fabrication. Since it has a fluorinated alky-chain configuration and is non-biodegradable, its treatment by conventional oxidation methods is typically very inefficient. In this work, novel three-phase fluidized bed reactor (3P-FBR, 7.5 cm in diameter, 50 cm high) that combines photo oxidation (UV/H2O2, one of AOPs (Advanced Oxidation Process) and adsorption (BT5 iron oxide as adsorbent) processes is designed for mineralizing and defluorinizing TFP wastewater. The experimental results reveal that TFP can be efficiently mineralized, and the BT5 that is circulated by aeration in the 3P-FBR system can remove the released fluoride ions in the reaction period. Irradiation with 254 nm UV and a 10 mM H2O2 dose yield a TOC removal of TFP (1.39 mM, equivalent to an initial TOC of 50 ppm) of over 99.95% in 2 h, and 99% of fluoride was removed by BT5 with an adsorption capacity of 24.1 mg-F g(-1). Copyright © 2013 Elsevier Ltd. All rights reserved.

Integral test of JENDL-3.3 on fast reactors

International Nuclear Information System (INIS)

Chiba, Gou; Hazama, Taira

2003-05-01

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

Reactor safety issues resolved by the 2D/3D program

International Nuclear Information System (INIS)

1995-09-01

The 2D/3D Program studied multidimensional thermal-hydraulics in a PWR core and primary system during the end-of-blowdown and post-blowdown phases of a large-break LOCA (LBLOCA), and during selected small-break LOCA (SBLOCA) transients. The program included tests at the Cylindrical Core Test Facility (CCTF), the Slab Core Test Facility (SCTF), and the Upper Plenum Test Facility (UPTF), and computer analyses using TRAC. Tests at CCTF investigated core thermal-hydraulics and overall system behavior while tests at SCTF concentrated on multidimensional core thermal-hydraulics. The UPTF tests investigated two-phase flow behavior in the downcomer, upper plenum, tie plate region, and primary loops. TRAC analyses evaluated thermal-hydraulic behavior throughout the primary system in tests as well as in PWRs. This report summarizes the test and analysis results in each of the main areas where improved information was obtained in the 2D/3D Program. The discussion is organized in terms of the reactor safety issues investigated. This report was prepared in a coordination among US, Germany and Japan. US and Germany have published the report as NUREG/IA-0127 and GRS-101 respectively. (author)

Reactor safety issues resolved by the 2D/3D program

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

The 2D/3D Program studied multidimensional thermal-hydraulics in a PWR core and primary system during the end-of-blowdown and post-blowdown phases of a large-break LOCA (LBLOCA), and during selected small-break LOCA (SBLOCA) transients. The program included tests at the Cylindrical Core Test Facility (CCTF), the Slab Core Test Facility (SCTF), and the Upper Plenum Test Facility (UPTF), and computer analyses using TRAC. Tests at CCTF investigated core thermal-hydraulics and overall system behavior while tests at SCTF concentrated on multidimensional core thermal-hydraulics. The UPTF tests investigated two-phase flow behavior in the downcomer, upper plenum, tie plate region, and primary loops. TRAC analyses evaluated thermal-hydraulic behavior throughout the primary system in tests as well as in PWRs. This report summarizes the test and analysis results in each of the main areas where improved information was obtained in the 2D/3D Program. The discussion is organized in terms of the reactor safety issues investigated. This report was prepared in a coordination among US, Germany and Japan. US and Germany have published the report as NUREG/IA-0127 and GRS-101 respectively. (author).

3D thermal-hydraulic analysis on core of PWR nuclear power station

International Nuclear Information System (INIS)

Yao Zhaohui; Wang Xuefang; Shen Mengyu

1997-01-01

Thermal hydraulic analysis of core is of great importance in reactor safety analysis. A computer code, thermal hydraulic analysis porous medium analysis (THAPMA), has been developed to simulate the flow and heat transfer characteristics of reactor components. It has been proved reliable by several numerical tests. In the THAPMA code, a new difference scheme and solution method have been studied in developing the computer software. For the difference scheme, a second order accurate, high resolution scheme, called WSUC scheme, has been proposed. This scheme is total variation bounded and unconditionally stable in convective numeral stability. Numerical tests show that the WSUC is better in accuracy and resolution than the 1-st order upwind, 2-nd order upwind, SOUCUP by Zhu and Rodi. In solution method, a modified PISO algorithm is used, which is not only simpler but also more accurate and more rapid in convergence than the original PISO algorithm. Moreover, the modified PISO algorithm can effectively solve steady and transient state problem. Besides, with the THAPMA code, the flow and heat transfer phenomena in reactor core have been numerically simulated in the light of the design condition of Qinshan PWR nuclear power station (the second-term project). The simulation results supply a theoretical basis for the core design

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

Apparatus and method for improving electrostatic precipitator performance by plasma reactor conversion of SO.sub.2 to SO.sub.3

Science.gov (United States)

Huang, Hann-Sheng; Gorski, Anthony J.

1999-01-01

An apparatus and process that utilize a low temperature nonequilibrium plasma reactor, for improving the particulate removal efficiency of an electrostatic precipitator (ESP) are disclosed. A portion of the flue gas, that contains a low level of SO.sub.2 O.sub.2 H.sub.2 O, and particulate matter, is passed through a low temperature plasma reactor, which defines a plasma volume, thereby oxidizing a portion of the SO.sub.2 present in the flue gas into SO.sub.3. An SO.sub.2 rich flue gas is thereby generated. The SO.sub.3 rich flue gas is then returned to the primary flow of the flue gas in the exhaust treatment system prior to the ESP. This allows the SO.sub.3 to react with water to form H.sub.2 SO.sub.4 that is in turn is absorbed by fly ash in the gas stream in order to improve the removal efficiency of the EPS.

Examination of U3Si2-Al fuel elements from the Oak Ridge Research Reactor

International Nuclear Information System (INIS)

Copeland, G.L.; Snelgrove, J.L.; Hofman, G.L.

1986-01-01

The results of postirradiation examination of low-enriched U 3 Si 2 fuel elements from the Oak Ridge Research Reactor are presented. The elements replaced standard high-enriched elements and were handled routinely except that the burnup of half the elements was extended beyond normal limits up to about 98% peak. The elements were manufactured by commercial fuel suppliers. The performance was completely satisfactory for all the elements

Enhanced CANDU 6 Reactor

International Nuclear Information System (INIS)

Azeez, S.; Alizadeh, A.; Girouard, P.

2005-01-01

Full text: The CANDU 6 power reactor is visionary in its approach, remarkable for its on-power refuelling capability and proven over years of safe, economical and reliable power production. Developed by Atomic Energy of Canada Ltd, the CANDU 6 design offers excellent performance utilizing state-of-the-art technology. The first CANDU 6 plants went into service in the early 1980's as leading edge technology and the design has been continuously advanced to maintain superior performance with an outstanding safety record. The first CANDU 6 plants- Gentilly 2 and Point Lepreau in Canada, Embalse in Argentina and Wolsong- Unit 1 in Korea have been in service for more than 21 years and are still producing electricity at peak performance and to the end of 2004, their average lifetime Capacity Factor was 83.2%. The newer CANDU 6 units in Romania (Cernavoda 1), Korea (Wolsong-Units 2, 3 and 4) and Qinshan (Phase III- Units 1 and 2) have also been performing at outstanding levels. The average lifetime Capacity Factor of the 10 CANDU 6 operating units around the world has been 87% to the end of 2004. Building on these successes, AECL is committed to the further development of this highly successful design, now focussing on meeting customer's needs for reduced costs, further improvements to plant operation and performance, enhanced safety and incorporating up-to-date technology as warranted. This has resulted in AECL embarking on improving the CANDU 6 design through an upgraded product termed as the 'Enhanced CANDU 6' (EC6)- which incorporates several attractive but proven features that will make the CANDU 6 reactor even more economical, safer and easier to operate. Some of the key features that will be incorporated in the EC6 include increasing the plant's power output, shortening the overall project schedule, decreasing the capital cost, dealing with obsolescence issues, optimizing maintenance outages and incorporating lessons learnt through feedback obtained from the

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.

Use of plate fuel elements for the RA3 reactor

International Nuclear Information System (INIS)

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

1992-01-01

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

Technical improvement for the output drive unit of the reactor protection system in QNPP

International Nuclear Information System (INIS)

Jiang Zuyue

1995-11-01

For improving the reliability of the output drive unit of the reactor protection system in Qinshan NPP, the former design of this part was improved and researched on the problem appeared during the commissioning and operation under the conditions of narrow process space of cabinets and unchanged overall arrangement: (1) The output relay modules was redesigned to unify the relay specification to improve the versatility, and also to improve the pin's contact by means of welding them directly on the printed circuit boards and to make the modules detachable by connectors instead of previously non-detachable. Th modules were connected in series by both power supply line and ground line which were finally connected at same point respectively, so that other protection signals can still be output correctly when a single module is removed. (2) The relay drive circuit was also redesigned for working in on-off state instead of in amplification to minimize the power consumption. On the other hand, the CMOS buffers were taken to couple the CMOS circuits to the TTL circuits. The actuating time for the new shutdown relay was decreased from the former 35 ms to 5 ms, the actuating time for the engineered safety feature drive signal relay was decreased from 10 ms to 6 ms after the above-mentioned improvements, the reliability of the RPS is remarkably improved and a great economic benefit is obtained. (4 refs., 3 figs., 2 tabs.)

Ethane dehydrogenation over nano-Cr{sub 2}O{sub 3} anode catalyst in proton ceramic fuel cell reactors to co-produce ethylene and electricity

Energy Technology Data Exchange (ETDEWEB)

Fu, Xian-Zhu; Luo, Xiao-Xiong; Luo, Jing-Li; Chuang, Karl T.; Sanger, Alan R. [Department of Chemical and Materials Engineering, University of Alberta, Edmonton, Alberta T6G2G6 (Canada); Krzywicki, Andrzej [NOVA Chemicals Corp., Calgary, Alberta T2P5C6 (Canada)

2011-02-01

Ethane and electrical power are co-generated in proton ceramic fuel cell reactors having Cr{sub 2}O{sub 3} nanoparticles as anode catalyst, BaCe{sub 0.8}Y{sub 0.15}Nd{sub 0.05}O{sub 3-{delta}} (BCYN) perovskite oxide as proton conducting ceramic electrolyte, and Pt as cathode catalyst. Cr{sub 2}O{sub 3} nanoparticles are synthesized by a combustion method. BaCe{sub 0.8}Y{sub 0.15}Nd{sub 0.05}O{sub 3-{delta}} (BCYN) perovskite oxides are obtained using a solid state reaction. The power density increases from 51 mW cm{sup -2} to 118 mW cm{sup -2} and the ethylene yield increases from about 8% to 31% when the operating temperature of the solid oxide fuel cell reactor increases from 650 C to 750 C. The fuel cell reactor and process are stable at 700 C for at least 48 h. Cr{sub 2}O{sub 3} anode catalyst exhibits much better coke resistance than Pt and Ni catalysts in ethane fuel atmosphere at 700 C. (author)

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

International Nuclear Information System (INIS)

Yagi, Masahiro; Sato, Mitsugu; Kakefuda, Kazuhiro

1999-01-01

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

Atomization of U3Si2 for research reactor fuel

International Nuclear Information System (INIS)

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

1995-01-01

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

Fissile fuel production and usage of thermal reactor waste fueled with UO2 by means of hybrid reactor system

International Nuclear Information System (INIS)

Ipek, O.

1997-01-01

The use of Fast Breeder Reactors to produce fissile fuel from nuclear waste and the operation of these reactors with a new neutron source are becoming today' topic. In the thermonuclear reactors, it is possible to use 2.45-14.1 MeV - neutrons which can be obtained by D-T, D-D Semicatalyzed (D-D) and other fusion reactions. To be able to do these, Hybrid Reactor System, which still has experimental and theoretical studies, have to be taken into consideration.In this study, neutronic analysis of hybrid blanket with grafit reflector, is performed. D-T driven fusion reaction is surrounded by UO 2 fuel layer and the production of ''2''3''9Pu fissile fuel from waste ''2''3''8U is analyzed. It is also compared to the other possible fusion reactions. The results show that 815.8 kg/year ''2''3''8Pu with D-T reaction and 1431.6 kg/year ''2''3''8Pu with semicatalyzed (D-D) reaction can be produced for 1000 MW fusion power. This means production of 2.8/ year and 4.94/ year LWR respectively. In addition, 1000 MW fusion flower is is multiplicated to 3415 MW and 4274 MW for D-T and semicatalyzed (D-D) reactions respectively. The system works subcritical and these values are 0.4115 and 0.312 in order. The calculations, ANISN-ORNL code, S 16 -P 3 approach and DLC36 data library are used

Final Report for the 1st Surveillance Test of the Reactor Pressure Vessel Material (CAPSULE 2) of Ulchin Nuclear Power Plant Unit 3

Energy Technology Data Exchange (ETDEWEB)

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

2006-12-15

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 1st surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejon after the capsule was transported from Ulchin 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 Ulchin unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsule 2 is 4.674E 18n/cm{sup 2}. The bias factor, the ratio of calculation/measurement, was 0.920 for the 1st testing and the calculational uncertainty,7.0% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 3.913E 18n/cm{sup 2} based on the end of 6th fuel cycle and it was predicted that the fluences of vessel inside surface at 16 and 32EFPY would reach 9.249E 18 and 1.834E 19n/cm{sup 2} based on the current calculation. The result through this analysis for Ulchin unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life.

Annex VII - Diagrams: 1. Reactor operation (1960-1977); 2. Mean daily reactor power density in 1977; 3. Monthly reactor power for 1977; 4. percent of utilization of experimental space in 1977; Prilog VII - Dijagrami: 1. Rad reaktora (MWh) po godinama (1960-1977); 2. Srednja dnevna snaga reaktora u 1977. godini; 3. Rad reaktora (MWh) po mesecima za 1977. godinu i 4. Procenat iskoriscenja eksperimentalnog prostora u 1977. godini

Energy Technology Data Exchange (ETDEWEB)

NONE

1977-12-15

This Annex includes the following diagrams: 1. Annual Reactor RA power production (MWh) for the period from 1960-1977; 2. Mean daily reactor power density MW in 1977; 3. Monthly reactor power production (MWh) for 1977; 4. percent of utilization of experimental space in 1977. [Serbo-Croat] Ovaj prilog sadrzi dijagrame: 1. Rad reaktora (MWh) po godinama (1960-1977); 2. Srednja dnevna snaga reaktora u 1977. godini; 3. Rad reaktora (MWh) po mesecima za 1977. godinu i 4. Procenat iskoriscenja eksperimentalnog prostora u 1977. godini.

Benchmark testing of CENDL-2 for U-fuel thermal reactors

International Nuclear Information System (INIS)

Zhang Baocheng; Liu Guisheng; Liu Ping

1995-01-01

Based on CENDL-2, NJOY-WIMS code system was used to generate 69-group constants, and do benchmark testing for TRX-1,2; BAPL-UO-2-1,2,3; ZEEP-1,2,3. All the results proved that CENDL-2 is reliable for thermal reactor calculations. (3 tabs.)

The 2nd reactor core of the NS Otto Hahn

International Nuclear Information System (INIS)

Manthey, H.J.; Kracht, H.

1979-01-01

Details of the design of the 2nd reactor core are given, followed by a brief report summarising the operating experience gained with this 2nd core, as well as by an evaluation of measured data and statements concerning the usefulness of the knowledge gained for the development of future reactor cores. Quite a number of these data have been used to improve the concept and thus the specifications for the fuel elements of the 3rd core of the reactor of the NS Otto Hahn. (orig./HP) [de

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

International Nuclear Information System (INIS)

Guidez, J.; Saturnin, A.

2016-01-01

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

Development of new ORIGEN2 data library sets for research reactors with light water cooled oxide and silicide LEU (20 w/o) fuels based on JENDL-3.3 nuclear data

International Nuclear Information System (INIS)

Liem, Peng Hong; Sembiring, Tagor Malem

2013-01-01

Highlights: • We developed new ORIGEN2 data library sets for research reactors based on JENDL-3.3. • The sets cover oxide and silicide LEU fuels with meat density up to 4.74 g U/cm 3 . • Two kinds of data library sets are available: fuel region and non-fuel regions. • We verified the new data library sets with other codes. • We validated the new data library against a non-destructive test. -- Abstract: New sets of ORIGEN2 data library dedicated to research/testing reactors with light water cooled oxide and silicide LEU fuel plates based on JENDL-3.3 nuclear data were developed, verified and validated. The new sets are considered to be an extension of the most recent release of ORIGEN2.2UPJ code, i.e. the ORLIBJ33 library sets. The newly generated ORIGEN2 data library sets cover both oxide and silicide LEU fuels with fuel meat density range from 2.96 to 4.74 g U/cm 3 used in the present and future operation of the Indonesian 30 MWth RSG GAS research reactor. The new sets are expected applicable also for other research/testing reactors which utilize similar fuels or have similar neutron spectral indices. In addition to the traditional ORIGEN2 library sets for fuel depletion analyses in fuel regions, in the new data library sets, new ORIGEN2 library sets for irradiation/activation analyses were also prepared which cover all representative non-fuel regions of RSG GAS such as reflector elements, irradiation facilities, etc. whose neutron spectra are significantly softer than fuel regions. Verification with other codes as well as validation with a non-destructive test result showed promising results where a good agreement was confirmed

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

Energy Technology Data Exchange (ETDEWEB)

Gallardo S, L F

1991-12-31

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

Effect of process temperature on morphology of CNTs grown in a vertically fluidized bed reactor with Fe{sub 2}O{sub 3}/Al{sub 2}O{sub 3} catalyst

Energy Technology Data Exchange (ETDEWEB)

Shukrullah, Shazia, E-mail: zshukrullah@gmail.com; Mohamed, Norani Muti, E-mail: noranimuti-mohamed@petronas.com.my; Shaharun, Maizatul Shima, E-mail: maizats@petronas.com.my [Department of Fundamental and Applied Sciences, Universiti Teknologi PETRONAS, 31750 Tronoh, Perak (Malaysia)

2015-07-22

Carbon nanotubes (CNTs) are one of the most researched materials due to their exceptional mechanical and electrical properties. Among the various techniques, catalytic chemical vapor deposition in a fluidized bed reactor is the most promising technique for bulk production of CNTs. To meet the demand of good quality along with the bulk production of CNTs, the effect of reaction temperature on the micro structures, morphology, diameter, quality and quantity of CNTs was investigated in these studies. CNTs were synthesized at process temperature ranging from 700-850°C by catalytic decomposition of C{sub 2}H{sub 4} on Fe{sub 2}O{sub 3}/Al{sub 2}O{sub 3} catalyst a vertical fluidized bed reactor. The microstructures of the grown CNTs at different reaction temperatures were investigated by using scanning electron microscope. The results of this study depicted a positive correlation between the average diameter of CNTs and reaction temperature. Narrow diameters (35∼40 nm) of CNTs with fewer defects were found at the low and mild temperatures, in particular 800°C. At this temperature, a dynamic equilibrium between the rate of C{sub 2}H{sub 4} decomposition and CNTs quantity was found due to maximum carbon diffusion over catalyst. The CNTs produced with Fe{sub 2}O{sub 3}/Al{sub 2}O{sub 3} catalyst wer e also exhibiting high quality with relatively small mean outer diameter and fewer surface defects.

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-09-15

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

Decontamination and decommissioning project of the TRIGA Mark-2 and 3 research reactors

Energy Technology Data Exchange (ETDEWEB)

Jung, K J; Baik, S T; Chung, U S; Jung, K H; Park, S K; Lee, B J; Kim, J K; Yang, S H

2000-01-01

During the review on the decommissioning plan and environmental impact assessment report by the KINS, the number of the inquired items were two hundred and fifty one, and the answers were made and sent until September 10, 1999, as the screened review results were reported to Ministry of Science and Technology(MOST) in December 14, 1999, all the reviews on the licence were over. Radioactive liquid wastes of 400 tons generated during the operation of the research reactors including reactor vessels are stored in the facility of the research reactor 1 and 2. Those liquid wastes have the low-level-radioactivity which can be discharged to the surroundings, but was wholly treated to be vaporized naturally by means of the increased numbers of the natural vaporization disposal facilities with the annual capacity of 200 tons for the purpose of the minimized environmental contamination.

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

International Nuclear Information System (INIS)

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

2013-01-01

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

D-3He fuel cycles for neutron lean reactors

International Nuclear Information System (INIS)

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

1989-01-01

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

Enhanced candu 6 reactor: status

International Nuclear Information System (INIS)

Azeez, S.; Girouard, P.

2006-01-01

The CANDU 6 power reactor is visionary in its approach, renowned for its on-power refuelling capability and proven over years of safe, economical and reliable power production. Developed by Atomic Energy of Canada Limited (AECL), the CANDU 6 design offers excellent performance utilizing state-of-the-art technology. The first CANDU 6 plants went into service in the early 1980s as leading edge technology and the design has been continuously advanced to maintain superior performance with an outstanding safety record. The first set of CANDU 6 plants - Gentilly 2 and Point Lepreau in Canada, Embalse in Argentina and Wolsong- Unit 1 in Korea - have been in service for more than 22 years and are still producing electricity at peak performance; to the end of 2004, their average Lifetime Capacity Factor was 83.2%. The newer CANDU 6 units in Romania (Cernavoda 1), Korea (Wolsong-Units 2, 3 and 4) and Qinshan (Phase III- Units 1 and 2) have also been performing at outstanding levels. The average lifetime Capacity Factor of the 10 CANDU 6 operating units around the world has been 87% to the end of 2004. Building on these successes, AECL is committed to the further development of this highly successful design, now focussing on meeting customers' needs for reduced costs, further improvements to plant operation and performance, enhanced safety and incorporating up-to-date technology, as warranted. This has resulted in AECL embarking on improving the CANDU 6 design through an upgraded product termed the ''Enhanced CANDU 6'' (EC6), which incorporates several attractive but proven features that make the CANDU 6 reactor even more economical, safer and easier to operate. Some of the key features that are being incorporated into the EC6 include increasing the plant's power output, shortening the overall project schedule, decreasing the capital cost, dealing with obsolescence issues, optimizing maintenance outages and incorporating lessons learnt through feedback obtained from the

Removal of FePO4 and Fe3(PO4)2 crystals on the surface of passive fillers in Fe0/GAC reactor using the acclimated bacteria

International Nuclear Information System (INIS)

Lai, Bo; Zhou, Yuexi; Yang, Ping; Wang, Juling; Yang, Jinghui; Li, Huiqiang

2012-01-01

Highlights: ► Fe 3 (PO 4 ) 2 and FePO 4 crystals would weaken treatment efficiency of Fe 0 /GAC reactor. ► Fe 3 (PO 4 ) 2 and FePO 4 crystals could be removed by the acclimated bacteria. ► FeS and sulfur in the passive film would be removed by the sulfur-oxidizing bacteria. ► Develop a cost-effective bio-regeneration technology for the passive fillers. - Abstract: As past studies presented, there is obvious defect that the fillers in the Fe 0 /GAC reactor begin to be passive after about 60 d continuous running, although the complicated, toxic and refractory ABS resin wastewater can be pretreated efficiently by the Fe 0 /GAC reactor. During the process, the Fe 3 (PO 4 ) 2 and FePO 4 crystals with high density in the passive film are formed by the reaction between PO 4 3− and Fe 2+ /Fe 3+ . Meanwhile, they obstruct the formation of macroscopic galvanic cells between Fe 0 and GAC, which will lower the wastewater treatment efficiency of Fe 0 /GAC reactor. In this study, in order to remove the Fe 3 (PO 4 ) 2 and FePO 4 crystals on the surface of the passive fillers, the bacteria were acclimated in the passive Fe 0 /GAC reactor. According to the results, it can be concluded that the Fe 3 (PO 4 ) 2 and FePO 4 crystals with high density in the passive film could be decomposed or removed by the joint action between the typical propionic acid type fermentation bacteria and sulfate reducing bacteria (SRB), whereas the PO 4 3− ions from the decomposition of the Fe 3 (PO 4 ) 2 and FePO 4 crystals were released into aqueous solution which would be discharged from the passive Fe 0 /GAC reactor. Furthermore, the remained FeS and sulfur (S) in the passive film also can be decomposed or removed easily by the oxidation of the sulfur-oxidizing bacteria. This study provides some theoretical references for the further study of a cost-effective bio-regeneration technology to solve the passive problems of the fillers in the zero-valent iron (ZVI) or Fe 0 /GAC reactor.

Safety assessments relating to the use of new fuels in research reactors: application to the case of FRM 2 reactor fuel

International Nuclear Information System (INIS)

Abou Yehia, H.; Bars, G.; Tran Dai

2001-01-01

After giving a brief reminder of the procedure applied in France for the licensing of the use of a new fuel type or design in a research reactor, we outline the main safety aspects associated with such a modification. Finally, by way of an example, we focus on the safety assessment relating to the IRIS irradiation device used in SILOE reactor, in particular for the qualification of the fuel dedicated to FRM II reactor of the Technical University of Munich. This qualification was carried out on a U 3 Si 2 fuel plate enriched to about 90 % in weight of 235 U and containing 1.5 g of uranium per cm 3 . The evaluation performed by the IPSN for GRS did not call into question the choice of U 3 Si 2 fuel plates for the FRM-II reactor. (authors)

Conclusions of the experts group of the RA reactor at the meeting held on November 2 and 3 1964 - Annex 12a; Prilog 12a - Zakljucci sa sastanka Strucnog kolegijuma reaktora RA odrzanog na dan 2. i 3. novembra 1964. godine

Energy Technology Data Exchange (ETDEWEB)

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

1964-12-15

Conclusions of the experts group of the RA reactor are related to: analyses of reactor operation at 6.5 MW power with heavy water coolant flow of 250 m{sup 3}/h (2 pumps rotation speed 1500 rotations/min); decisions of future operation; further preparation activities related to reactor operation in forced regime and reduced cooling conditions.

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

International Nuclear Information System (INIS)

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

2002-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-04-01

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

Apollo-L2, an advanced fuel tokamak reactor utilizing direct conversion

International Nuclear Information System (INIS)

Emmert, G.A.; Kulcinski, G.L.; Blanchard, J.P.; El-Guebaly, L.A.; Khater, H.Y.; Santarius, J.F.; Sawan, M.E.; Sviatoslavsky, I.N.; Wittenberg, L.J.; Witt, R.J.

1989-01-01

A scoping study of a tokamak reactor fueled by a D- 3 He plasma is presented. The Apollo D- 3 He tokamak capitalizes on recent advances in high field magnets (20 T) and utilizes rectennas to convert the synchrotron radiation directly to electricity. The low neutron wall loading (0.1 MW/m 2 ) permits a first wall lasting the life of the plant and enables the reactor to be classified as inherently safe. The cost of electricity is less than that from a similar power level DT reactor. 10 refs., 1 fig., 4 tabs

Uranium-fuel thermal reactor benchmark testing of CENDL-3

International Nuclear Information System (INIS)

Liu Ping

2001-01-01

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

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

International Nuclear Information System (INIS)

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

1991-07-01

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

Advanced passive PWR AC-600: Development orientation of nuclear power reactors in China for the next century

International Nuclear Information System (INIS)

Huang Xueqing; Zhang Senru

1999-01-01

Based on Qinshan II Nuclear Power Plant that is designed and constructed by way of self-reliance, China has developed advanced passive PWR AC-600. The design concept of AC-600 not only takes the real situation of China into consideration, but also follows the developing trend of nuclear power in the world. The design of AC-600 has the following technical characteristics: Advanced reactor: 18-24 month fuel cycle, low neutron leakage, low power density of the core, no any penetration in the RPV below the level of the reactor coolant nozzles; Passive safety systems: passive emergency residual heat removal system, passive-active safety injection system, passive containment cooling system and main control room habitability system; System simplified and the number of components reduced; Digital I and C; Modular construction. AC-600 inherits the proven technology China has mastered and used in Qirtshan 11, and absorbs advanced international design concepts, but it also has a distinctive characteristic of bringing forth new ideas independently. It is suited to Chinese conditions and therefore is expected to become an orientation of nuclear power development by self-reliance in China for the next century. (author)

PR-EDB: Power Reactor Embrittlement Database Version 3

International Nuclear Information System (INIS)

Wang, Jy-An John; Subramani, Ranjit

2008-01-01

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

PR-EDB: Power Reactor Embrittlement Database - Version 3

Energy Technology Data Exchange (ETDEWEB)

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

2008-03-01

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

Removal in a lump of JRR-3 nuclear reactor

International Nuclear Information System (INIS)

Ohnishi, Nobuaki; Suzuki, Masanori; Nagase, Tetsuo; Watanabe, Morinari.

1989-01-01

The research reactor JRR-3 in Japan Atomic Energy Research Institute is called 'Home made No.1 reactor' as all except fuel and heavy water as the moderator and coolant were manufactured in Japan. The JRR-3 attained the criticality in 1962, and the cumulative time of operation reached 47135.5 hours, and the cumulative power output reached 419073.5 MWh. It was stopped in 1983. During the period, it was utilized for beam experiment, irradiation of fuel and materials, RI production and others. In order to cope with the expansion of utilization and the advance of utilizing technology of the research reactor, the reconstruction works are in progress, and the criticality of the reconstructed reactor is expected in 1990. On the site where the old reactor is removed, the reactor of different type is installed, and the first large cold neutron source is equipped. In this report, as to the removal of the old reactor proper, the method of working and the results are described. Considering the period of working, the cost and the management of the removed reactor, in the case of the JRR-3, the method of carrying it out in a lump was adopted as the optimum removal method. The plan, procedure and results of the removal working are reported. (K.I.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-10-01

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

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

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

International Nuclear Information System (INIS)

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

1983-07-01

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

3D simulation of CANDU reactor regulating system

International Nuclear Information System (INIS)

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

2013-01-01

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

Analyses for MARIA Research Reactor with RELAP/MOD3 code

International Nuclear Information System (INIS)

Szczurek, J.; Czerski, P.

2004-01-01

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

BR2 Reactor: Irradiation of fuels

International Nuclear Information System (INIS)

Verwimp, A.

2005-01-01

Safe, reliable and economical operation of reactor fuels, both UO 2 and MOX types, requires in-pile testing and qualification up to high target burn-up levels. In-pile testing of advanced fuels for improved performance is also mandatory. The objectives of research performed at SCK-CEN are to perform Neutron irradiation of LWR (Light Water Reactor) fuels in the BR2 reactor under relevant operating and monitoring conditions, as specified by the experimenter's requirements and to improve the on-line measurements on the fuel rods themselves

Death analysis of residents in an area of twenty kilometers around Qinshan nuclear power station

International Nuclear Information System (INIS)

Ma Mingqiang; Lu Zhunrong; Zheng Wen; Sun Peizhi

2001-01-01

Objective: To set up a data bank for residents health condition 20 kilometers within around Qinshan nuclear power station. Methods: Combining with retrospective investigation, the relevant data were acquired from medical certification for resident's death reported by all local disease surveillance. Results: The mortality rate of these residents from 1988 to 1999 was 6.92%. The first course of mortality was diseases of respiratory system, the second of circulatory system, and the third was malignant tumor. The first 5 death causes among all male and female persons were diseases of respiratory system and circulatory system, malignant tumor, injuries and poisoning, diseases of digestive system. The mortality rate for malignant tumor was 121.33/100000 (the standard death rate is : 100.13/100000), and liver cancer was the first death cause, while lung and stomach cancers, the second and the third, respectively. The main death causes in juvenile and youth was leukemia, but liver cancer and lung cancer were the main courses of death in the middle-aged, and in old people, lung and liver cancers. Conclusion: The chronic non-infectious diseases in respiratory system, circulatory system etc, are the major death causes in the residents, the mortality rate for malignant tumor in them is lower than that reported by provincial disease surveillance station

The use of U3Si2 dispersed in aluminum in plate-type fuel elements for research and test reactors

International Nuclear Information System (INIS)

Snelgrove, J.L.; Domagala, R.F.; Hofman, G.L.; Wiencek, T.C.; Copeland, G.L.; Hobbs, R.W.; Senn, R.L.

1987-10-01

A high-density fuel based on U 3 Si 2 dispersed in aluminum has been developed and tested for use in converting plate-type research and test reactors from the use of highly enriched uranium to the use of low-enriched uranium. Results of preirradiation testing and the irradiation and postirradiation examination of miniature fuel plates and full-sized fuel elements are summarized. Swelling of the U 3 Si 2 fuel particles is a linear function of the fission density in the particle to well beyond the fission density achievable in low-enriched fuels. U 3 Si 2 particle swelling rate is approximately the same as that of the commonly used UAl/sub x/ fuel particle. The presence of minor amounts of U 3 Si or uranium solid solution in the fuel result in greater, but still acceptable, fuel swelling. Blister threshold temperatures are at least as high as those of currently used fuels. An exothermic reaction occurs near the aluminum melting temperature, but the measured energy releases were low enough not to substantially worsen the consequences of an accident. U 3 Si 2 -aluminum dispersion fuel with uranium densities up to at least 4.8 Mg/m 3 is a suitable LEU fuel for typical plate-type research and test reactors. 42 refs., 28 figs., 7 tabs

TARMS, an on-line boiling water reactor operation management system. [3 D core simulator LOGOS 2

Energy Technology Data Exchange (ETDEWEB)

Iwamoto, T.; Sakurai, S.; Uematsu, H.; Tsuiki, M.; Makino, K.

1984-12-01

The TARMS (Toshiba Advanced Reactor Management System) software package was developed as an effective on-line, on-site tool for boiling water reactor core operation management. It was designed to support a complete function set to meet the requirement to the current on-line process computers. The functions can be divided into two categories. One is monitoring of the present core power distribution as well as related limiting parameters. The other is aiding site engineers or reactor operators in making the future reactor operating plan. TARMS performs these functions with a three-dimensional BWR core physics simulator LOGOS 2, which is based on modified one-group, coarse-mesh nodal diffusion theory. A method was developed to obtain highly accurate nodal powers by coupling LOGOS 2 calculations with the readings of an in-core neutron flux monitor. A sort of automated machine-learning method also was developed to minimize the errors caused by insufficiency of the physics model adopted in LOGOS 2. In addition to these fundamental calculational methods, a number of core operation planning aid packages were developed and installed in TARMS, which were designed to make the operator's inputs simple and easy.

Completion of reconstruction for Japan Research Reactor No.3

International Nuclear Information System (INIS)

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

1992-01-01

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

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

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

International Nuclear Information System (INIS)

Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi

2000-06-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-06-01

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

Sterilization of E. coli bacterium in a flowing N2-O2 post-discharge reactor

International Nuclear Information System (INIS)

Villeger, S; Cousty, S; Ricard, A; Sixou, M

2003-01-01

Effective destruction of Escherichia coli (E. coli) bacteria has been obtained in a flowing N 2 -O 2 microwave post-discharge reactor. The sterilizing agents are the O atoms and the UV emissions of NOβ which are produced by N and O atoms recombination in the reactor. In the following plasma conditions: pressure 5 Torr, flow rate 1 L n min -1 , microwave power of 100 W in a quartz tube of 5 mm, an O atom density of 2.5x10 15 cm -3 is measured by NO titration in the post-discharge reactor with UV emission in a N 2 -(5%)O 2 gas mixture. Full destruction of 10 13 cfu ml -1 E. coli is observed after a treatment time of 25 min. (rapid communication)

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)

Course of operators of the RA-3 reactor

International Nuclear Information System (INIS)

Caligiuri, G.A.

1983-01-01

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

Decommissioning of reactor facilities (2). Required technology

International Nuclear Information System (INIS)

Yanagihara, Satoshi

2014-01-01

Decommissioning of reactor facilities was planned to perform progressive dismantling, decontamination and radioactive waste disposal with combination of required technology in a safe and economic way. This article outlined required technology for decommissioning as follows: (1) evaluation of kinds and amounts of residual radioactivity of reactor facilities with calculation and measurement, (2) decontamination technology of metal components and concrete structures so as to reduce worker's exposure and production of radioactive wastes during dismantling, (3) dismantling technology of metal components and concrete structures such as plasma arc cutting, band saw cutting and controlled demolition with mostly remote control operation, (3) radioactive waste disposal for volume reduction and reuse, and (4) project management of decommissioning for safe and rational work to secure reduction of worker's exposure and prevent the spreading of contamination. (T. Tanaka)

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-15

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

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

International Nuclear Information System (INIS)

Anon.

1976-01-01

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

OTUS - Reactor inventory management system based on ORIGEN2

Energy Technology Data Exchange (ETDEWEB)

Poellaenen, R; Toivonen, H; Lahtinen, J; Ilander, T

1995-10-01

ORIGEN2 is a computer code that calculates nuclide composition and other characteristics of nuclear fuel. The use of ORIGEN2 requires good knowledge in reactor physics. However, once the input has been defined for a particular reactor type, the calculations can be easily repeated for any burnup and decay time. This procedure produces large output files that are difficult to handle manually. A new computer code, known as OTUS, was designed to facilitate the postprocessing of the data. OTUS makes use of the inventory files precalculated with ORIGEN2 in a way that enables their versatile treatment for different safety analysis purposes. A data base is created containing a comprehensive set of ORIGEN2 calculations as a function of fuel burnup and decay time. OTUS is a reactor inventory management system for a microcomputer with Windows interface. Four major data operations are available: (1) Build data modifies ORIGEN2 output data into a suitable format, (2) View data enables flexible presentation of the data as such, (3) Different calculations, such as nuclide ratios and hot particle characteristics, can be performed for severe accident analyses, consequence analyses and research purposes, (4) Summary files contain both burnup dependent and decay time dependent inventory information related to the nuclide and the reactor specified. These files can be used for safeguards, radiation monitoring and safety assessment. (orig.) (22 refs., 29 figs.).

An introduction to the design, commissioning and operation of nuclear air cleaning systems for Qinshan Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Xinliang Chen; Jiangang Qu; Minqi Shi [Shanghai Nuclear Engineering Research and Design Institute (China)] [and others

1995-02-01

This paper introduces the design evolution, system schemes and design and construction of main nuclear air cleaning components such as HEPA filter, charcoal adsorber and concrete housing etc. for Qinshan 300MW PWR Nuclear Power Plant (QNPP), the first indigenously designed and constructed nuclear power plant in China. The field test results and in-service test results, since the air cleaning systems were put into operation 18 months ago, are presented and evaluated. These results demonstrate that the design and construction of the air cleaning systems and equipment manufacturing for QNPP are successful and the American codes and standards invoked in design, construction and testing of nuclear air cleaning systems for QNPP are applicable in China. The paper explains that the leakage rate of concrete air cleaning housings can also be assured if sealing measures are taken properly and embedded parts are designed carefully in the penetration areas of the housing and that the uniformity of the airflow distribution upstream the HEPA filters can be achieved generally no matter how inlet and outlet ducts of air cleaning unit are arranged.

Atomic Energy of Canada Limited annual report 2000-2001

International Nuclear Information System (INIS)

2001-01-01

This is the annual report of the Atomic Energy of Canada Limited for the year ending March 31, 2001 and summarizes the activities of AECL during the period 2000-2001. The activities covered in this report include the CANDU reactor business, with progress being reported in the construction of two CANDU 6 reactors for the Qinshan CANDU project in China, the anticipated completion of Cernavoda unit 2, the completion of spent fuel storage at Cernavoda unit 1 in Romania, as well as the service business with New Brunswick Power, Ontario Power Generation, Bruce Power and Hydro Quebec in the refurbishment of operating, CANDU reactors. In the R and D programs discussions continue on funding for the Canadian Neutron Facility for Materials Research (CNF) and progress on the Maple medical isotope reactor

Atomic Energy of Canada Limited annual report 2000-2001

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

This is the annual report of the Atomic Energy of Canada Limited for the year ending March 31, 2001 and summarizes the activities of AECL during the period 2000-2001. The activities covered in this report include the CANDU reactor business, with progress being reported in the construction of two CANDU 6 reactors for the Qinshan CANDU project in China, the anticipated completion of Cernavoda unit 2, the completion of spent fuel storage at Cernavoda unit 1 in Romania, as well as the service business with New Brunswick Power, Ontario Power Generation, Bruce Power and Hydro Quebec in the refurbishment of operating, CANDU reactors. In the R and D programs discussions continue on funding for the Canadian Neutron Facility for Materials Research (CNF) and progress on the Maple medical isotope reactor.

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.)

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

2D and 3D CFD modelling of a reactive turbulent flow in a double shell supercritical water oxidation reactor

International Nuclear Information System (INIS)

Moussiere, S.; Roubaud, A.; Fournel, B.; Joussot-Dubien, C.; Boutin, O.; Guichardon, P.

2012-01-01

In order to design and define appropriate dimensions for a supercritical oxidation reactor, a comparative 2D and 3D simulation of the fluid dynamics and heat transfer during an oxidation process has been performed. The solver used is a commercial code, Fluent 6.2 (R). The turbulent flow field in the reactor, created by the stirrer, is taken into account with a k-omega model and a swirl imposed to the fluid. In the 3D case the rotation of the stirrer can be modelled using the sliding mesh model and the moving reference frame model. This work allows comparing 2D and 3D velocity and heat transfer calculations. The predicted values (mainly species concentrations and temperature profiles) are of the same order in both cases. The reactivity of the system is taken into account with a classical Eddy Dissipation Concept combustion model. Comparisons with experimental temperature measurements validate the ability of the CFD modelling to simulate the supercritical water oxidation reactive medium. Results indicate that the flow can be considered as plug flow-like and that heat transfer is strongly enhanced by the stirring. (authors)

Distribution of energy of impulses of the modernized IBR-2 REACTOR

International Nuclear Information System (INIS)

Tayibov, L.A; Mehtiyeva, R.N.; )

2011-01-01

Full text: For the modernized IBR-2 reactor there are two main reasons causing fluctuations of energy of impulses [1,3] on low power of stochastic fluctuations, on the nominal - giving rise to fluctuations of external reactance. The fluctuations of pulse energy is quite significant (20%). They affect the dynamics of the reactor, the process of regulation, starting, as well as the work of the experimental apparatus, etc. It is clear that research of fluctuation of energy of impulses has special value for the IBR-2 type reactor. Sufficient information about the statistical properties of the reactor noise gives the density distribution of the energy pulse power. We used the usual procedure of statistical analysis of time series. Calculated pulse energy of density and the parameters of this distribution.

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

International Nuclear Information System (INIS)

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

1977-01-01

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

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

International Nuclear Information System (INIS)

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

1977-01-01

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

Reactivity feedback coefficients of a material test research reactor fueled with high-density U{sub 3}Si{sub 2} dispersion fuels

Energy Technology Data Exchange (ETDEWEB)

Muhammad, Farhan [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Nilore, Islamabad 45650 (Pakistan)], E-mail: farhan73@hotmail.com; Majid, Asad [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Nilore, Islamabad 45650 (Pakistan)

2008-10-15

The reactivity feedback coefficients of a material test research reactor fueled with high-density U{sub 3}Si{sub 2} dispersion fuels were calculated. For this purpose, the low-density LEU fuel of an MTR was replaced with high-density U{sub 3}Si{sub 2} LEU fuels currently being developed under the RERTR program. Calculations were carried out to find the fuel temperature reactivity coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the average values of fuel temperature reactivity feedback coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient from 20 deg. C to 100 deg. C, at the beginning of life, followed the relationships (in units of {delta}k/k x 10{sup -5} K{sup -1}) -2.116 - 0.118 {rho}{sub U}, 0.713 - 37.309/{rho}{sub U} and -12.765 - 34.309/{rho}{sub U}, respectively for 4.0 {<=} {rho}{sub U} (g/cm{sup 3}) {<=} 6.0.

The qualification of U3O8 as research reactor fuel

International Nuclear Information System (INIS)

Krull, W.

1983-01-01

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

Benchmark calculations for VENUS-2 MOX -fueled reactor dosimetry

International Nuclear Information System (INIS)

Kim, Jong Kung; Kim, Hong Chul; Shin, Chang Ho; Han, Chi Young; Na, Byung Chan

2004-01-01

As a part of a Nuclear Energy Agency (NEA) Project, it was pursued the benchmark for dosimetry calculation of the VENUS-2 MOX-fueled reactor. In this benchmark, the goal is to test the current state-of-the-art computational methods of calculating neutron flux to reactor components against the measured data of the VENUS-2 MOX-fuelled critical experiments. The measured data to be used for this benchmark are the equivalent fission fluxes which are the reaction rates divided by the U 235 fission spectrum averaged cross-section of the corresponding dosimeter. The present benchmark is, therefore, defined to calculate reaction rates and corresponding equivalent fission fluxes measured on the core-mid plane at specific positions outside the core of the VENUS-2 MOX-fuelled reactor. This is a follow-up exercise to the previously completed UO 2 -fuelled VENUS-1 two-dimensional and VENUS-3 three-dimensional exercises. The use of MOX fuel in LWRs presents different neutron characteristics and this is the main interest of the current benchmark compared to the previous ones

Safety demonstration test (SR-3/S1C-3/S2C-3/SF-2) plan using the HTTR. Contract research

International Nuclear Information System (INIS)

Nakagawa, Shigeaki; Sakaba, Nariaki; Takamatsu, Kuniyoshi; Takada, Eiji; Tochio, Daisuke; Ohwada, Hiroyuki

2005-03-01

Safety demonstration tests using the HTTR are to be conducted from the FY2002 to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to not only the commercial HTGRs but also the research and development for the VHTR that is one of the Generation IV reactor candidates. This paper describes the reactivity insertion test (SR-3), the coolant flow reduction test by tripping of gas circulators (S1C-3, S2C-3), and the partial flow loss of coolant test (SF-2) planned in March 2005 with their detailed test method, procedure and results of pre-test analysis. From the analytical results, it was verified that the negative reactivity feedback effect of the core brings the reactor power safely to a stable level without a reactor scram. (author)

BR2 reactor neutron beams

International Nuclear Information System (INIS)

Neve de Mevergnies, M.

1977-01-01

The use of reactor neutron beams is becoming increasingly more widespread for the study of some properties of condensed matter. It is mainly due to the unique properties of the ''thermal'' neutrons as regards wavelength, energy, magnetic moment and overall favorable ratio of scattering to absorption cross-sections. Besides these fundamental reasons, the impetus for using neutrons is also due to the existence of powerful research reactors (such as BR2) built mainly for nuclear engineering programs, but where a number of intense neutron beams are available at marginal cost. A brief introduction to the production of suitable neutron beams from a reactor is given. (author)

Fission product release from SLOWPOKE-2 reactors

Energy Technology Data Exchange (ETDEWEB)

Harnden-Gillis, A M.C. [Queen` s Univ., Kingston, ON (Canada). Dept. of Physics

1994-12-31

Increasing radiation fields at several SLOWPOKE-2 reactors fuelled with highly enriched uranium aluminum alloy fuel have begun to interfere with the daily operation of these reactors. To investigate this phenomenon, samples of reactor container water and gas from the headspace were obtained at four SLOWPOKE-2 reactor facilities and examined by gamma ray spectroscopy methods. These radiation fields are due to the circulation of fission products within the reactor container vessel. The most likely source of the fission product release is an area of uranium-bearing material exposed to the coolant at the end weld line which originated at the time of fuel fabrication. The results of this study are compared with observations from an underwater visual examination of one core and the metallographic examination of archived fuel elements. 19 refs., 4 tabs., 8 figs.

Installation and commissioning of operation nuclear power plant reactor protection system modernization project

International Nuclear Information System (INIS)

Lu Weiwei

2010-01-01

Qinshan Nuclear Power Plant is the first nuclear power plant in mainland China; it is also the first one which realizes the modernization of analog technology based Reactor Protection System in the operation nuclear power plant of China. The implementation schedule is the shortest one which use same digital technology platform (TELEPERM XS of AREVA NP) to modifying the safety class I and C system in the world, the whole project spent 28 months from equipment contract signed to putting system into operation. It open up a era for operation nuclear power plant using mature digital technology to make safety class I and C system modernization in China. The important practical significance of this successful project is very obvious. This article focus on two important project stage--equipment installation and system commissioning, it is based on a large number of engineering implementation fact, it covers the problems and solutions happened during the installation and commission. The purpose of the article is to share the experience and lessons of safety I and C system modernization for other operation nuclear power plant. (authors)

FORE-2, Thermohydraulics and Space-Independent Reactor Kinetics for Transients

International Nuclear Information System (INIS)

Fox, J.N.; Lawler, B.E.; Butz, H.R.; Heames, T.J.

1984-01-01

1 - Description of problem or function: FORE2 is a coupled thermal hydraulics-point kinetics digital computer code designed to calculate significant reactor parameters under steady-state conditions, or as functions of time during transients. The transients may result from a programmed reactivity insertion or a power change. Variable inlet coolant flow rate and temperature are considered. The code calculates the reactor power, the individual reactivity feedbacks, and the temperature of coolant, cladding, fuel, structure, and additional material for up to seven axial positions in three channel types which represent radial zones of the reactor. The heat of fusion, accompanying fuel melting, the liquid metal voiding reactivity, and the spatial and the time variation of the fuel cladding gap coefficient due to changes in gap size are considered. 2 - Method of solution: FORE2 input consists of property data, geometry, power and flow distribution factors, external time varying functions, experimental coefficients, and termination data. The differential equations for fluid flow, heat transfer, and point neutronics are solved by explicit finite-difference procedures. 3 - Restrictions on the complexity of the problem: Reactor excursions which can be calculated are restricted to those transients in which the reactor is not substantially destroyed. As a general rule, changes in reactor geometry and composition during an excursion are limited to those cases in which the reactivity effects of the changes may be considered as small perturbations of the initial system. Thus, accidents involving large-scale disassembly and bulk meltdown of a core are not covered by FORE2. FORE2 is valid only while the core retains its initial geometry

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

Science.gov (United States)

Longoni, Gianluca; Anderson, Stanwood L.

2009-08-01

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

3 Investment Scenarios for Fast Reactors

International Nuclear Information System (INIS)

Shoai Tehrani, Bianka; Da Costa, Pascal

2013-01-01

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

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

International Nuclear Information System (INIS)

Shi, Dunfu

2015-01-01

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

Pressurized water reactor simulator. Workshop material. 2. ed

International Nuclear Information System (INIS)

2005-01-01

The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development. And the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21, 2nd edition, 'WWER-1000 Reactor Simulator' (2005). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23, 2nd edition, 'Boiling Water Reactor Simulator' (2005). This report consists of course material for workshops using a pressurized water reactor simulator

Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

Energy Technology Data Exchange (ETDEWEB)

Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

2016-01-01

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.

Design and computational analysis of passive siphon breaker for 49-2 swimming pool reactor

International Nuclear Information System (INIS)

Yue Zhiting; Song Yunpeng; Liu Xingmin; Zou Yao; Wu Yuanyuan

2014-01-01

Based on safety considerations, a passive siphon breaker will be added to the primary cooling system of 49-2 Swimming Pool Reactor (SPR). With the breaker location determined, the capability of siphon breakers with diameters of 1.5 cm and 2.0 cm was calculated and analyzed respectively by RELAP5/MOD3.3 code. The results show that in the condition of large break loss of coolant accident these two sizes of siphon breakers are able to break the siphon phenomena, and maintain the pool water level above the reactor core when the reactor and the pump are shutdown. In the end, to be conservative, the siphon breaker with diameter of 2.0 cm is adopted. (authors)

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

International Nuclear Information System (INIS)

Vicens, Hugo E.; Quintana, Jorge A.

2001-01-01

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Manpower development for safe operation of nuclear power plant. China. Steam generator maintenance, cleaning and repair. Activity: 3.1.8-Task-04. Technical report

International Nuclear Information System (INIS)

Esposito, J.N.

1994-01-01

The objective of this mission was to present detailed state-of-the-art information on steam generator design, operations and maintenance, to the management, engineers and operators of the Qinshan Nuclear Power Plants. In addition, some limited operation was presented by the Qinshan Nuclear Power Plant representatives in order to aid in focussing the presentations and promoting a high level of discussion

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

International Nuclear Information System (INIS)

Duerigen, S.; Fridman, E.

2011-01-01

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

Effect of Promoter Concentration on CO2 Separation Using K2CO3 With Reactive Absorption Method in Reactor Packed Column

Directory of Open Access Journals (Sweden)

Monde Junety

2018-01-01

Full Text Available The presence of carbon dioxide (CO2 in the gas is not expected because CO2 can reduce heating value and CO2 is the major emission contributor into the atmosphere. Various separation technologies can be used to reduce CO2 content and improve quality of gas. Chemical or reactive absorption is most widely used because it provides higher removal rate. This paper will study the effect of the addition di ethanolamine (DEA concentration into aqueous 30wt.% potassium carbonate(K2CO3 with reactive absorption method in a reactor packed column at temperature from 40°C to 80°C, DEA concentration range of (1% - 3% and absorbent flow rate (0.5, 0.75 and 1 L. min1. Contacting the gas and absorbent are countercurrent flow in packed column with 1.5 m high and 50 mm in diameter. The absorption column was randomly packed with a packing material raschig rings 5 mm in diameter. The CO2 loading in the liquid samples was determined by titration. It is found that the best result of CO2 loading is 0.065594 mole/mole K2CO3 and CO2 removal 28%. The result show that the loading capacity (mole CO2/mole K2CO3 and CO2 removal increased with the increase of DEA concentration.

3D computer visualization and animation of CANDU reactor core

International Nuclear Information System (INIS)

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

1999-01-01

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

Experiment for search for sterile neutrino at SM-3 reactor

Science.gov (United States)

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

2016-11-01

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

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

KAUST Repository

Karsenty, Florent

2012-08-16

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

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.

Planned Scientific programs around the Triga Mark 2 Reactor

International Nuclear Information System (INIS)

Majah, M Ibn.

2007-01-01

Full text: Nuclear techniques have been introduced to Morocco since the sixties. After the energy crisis of 1973, Morocco decides to create the National Center for Energy Sciences and Nuclear Techniques (CNESTEN) under the supervision of the Ministry of high Education and Research, with a research commercial and support vocation. CNESTEN is in charge of promoting nuclear application, to act as technical support for the authorities and to prepare the technological basis for nuclear power option. In 1998, CNESTEN started the construction of Nuclear Research Centre. The on going activities cover many sectors : earth and environmental sciences, high energy physics, safety and security, waste management. In 2001, CNESTEN started the construction of a 2MW TRiga Mark 2 Reactor, with the possibility to increase the power to 3 MW. The construction was achieved in January 2007. The operation of the reactor is expected for April 2007. The program of the utilization of the reactor was established with th contribution of the university and with the assistance of IAEA. Some of the experimental set-up installed around the reactor have been designed. CNESTEN has developed cooperation with Nuclear research centres from other countries and is receiving visitors and trainees mainly through the IAEA [fr

Validation of the MC{sup 2}-3/DIF3D Code System for Control Rod Worth via the BFS-75-1 Reactor Physics Experiment

Energy Technology Data Exchange (ETDEWEB)

Yun, Sunghwan; Kim, Sang Ji [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In this paper, control rod worths of the BFS-75-1 reactor physics experiments were examined using continuous energy MCNP models and deterministic MC2-3/DIF3D models based on the ENDF/B-VII.0 library. We can conclude that the ENDF/B-VII.0 library shows very good agreement in small-size metal uranium fuel loaded core which is surrounded by the depleted uranium blanket. However, the control rod heterogeneity effect reported by the reference is not significant in this problem because the tested control rod models were configured by single rod. Hence comparison with other control rod worth measurements data such as the BFS-109-2A reactor physics experiment is planned as a future study. The BFS-75-1 critical experiment was carried out in the BFS-1 facility of IPPE in Russia within the framework of validating an early phase of KALIMER- 150 design. The Monte-Carlo model of the BFS- 75-1 critical experiment had been developed. However, due to incomplete information for the BFS- 75-1 experiments, Monte-Carlo models had been generated for the reference criticality and sodium void reactivity measurements with disk-wise homogeneous model. Recently, KAERI performed another physics experiment, BFS-109-2A, by collaborating with Russian IPPE. During the review process of the experimental report of the BFS-109-2A critical experiments, valuable information for the BFS-1 facility which can also be used for the BFS-75-1 experiments was discovered.

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)

Effect of Co3O4 and CeO2 Infiltration on the Activity of a LSM15/GDC10 Highly Porous Electrochemical Reactor

DEFF Research Database (Denmark)

Ippolito, Davide; Kammer Hansen, Kent

2014-01-01

The reduction of air pollution has become an international concern over the last ten years because of increases in emissions from mobile and stationary sources. Among these sources, volatile organic compounds (VOC) represent a serious environmental problem, together with NOx, SOx and particulate...... VOC component of Diesel engine exhausts, over a wide range of temperatures. The entire reactor was thought as a highly porous catalytic filter for a possible application in a Diesel exhausts purification system. The porous reactor was used as a backbone for the infiltration of Co3O4 and Co3O4/CeO2...

Reactor building integrity testing: A novel approach at Gentilly 2 - principles and methodology

International Nuclear Information System (INIS)

Collins, N.; Lafreniere, P.

1991-01-01

In 1987, Hydro-Quebec embarked on an ambitious development program to provide the Gentilly 2 nuclear power station with an effective, yet practical reactor building Integrity Test. The Gentilly 2 Integrity Test employs an innovative approach based on the reference volume concept. It is identified as the Temperature Compensation Method (TCM) System. This configuration has been demonstrated at both high and low test pressure and has achieved extraordinary precision in the leak rate measurement. The Gentilly 2 design allows the Integrity Test to be performed at a nominal 3 kPa(g) test pressure during an (11) hour period with the reactor at full power. The reactor building Pressure Test by comparison, is typically performed at high pressure 124 kPa(g)) in a 7 day window during an annual outage. The Integrity Test was developed with the goal of demonstrating containment availability. Specifically it was purported to detect a leak or hole in the 'bottled-up' reactor building greater in magnitude than an equivalent pipe of 25 mm diameter. However it is considered feasible that the high precision of the Gentilly 2 TCM System Integrity Test and a stable reactor building leak characteristic will constitute sufficient grounds for the reduction of the Pressure Test frequency. It is noted that only the TCM System has, to this date, allowed a relevant determination of the reactor building leak rate at a nominal test pressure of 3 kPa(g). Classical method tests at low pressure have lead to inconclusive results due to the high lack of precision

Keeping research reactors relevant: A pro-active approach for SLOWPOKE-2

International Nuclear Information System (INIS)

Cosby, L.R.; Bennett, L.G.I.; Nielsen, K.; Weir, R.

2010-01-01

The SLOWPOKE is a small, inherently safe, pool-type research reactor that was engineered and marketed by Atomic Energy of Canada Limited (AECL) in the 1970s and 80s. The original reactor, SLOWPOKE-1, was moved from Chalk River to the University of Toronto in 1970 and was operated until upgraded to the SLOWPOKE-2 reactor in 1973. In all, eight reactors in the two versions were produced and five are still in operation today, three having been decommissioned. All of the remaining reactors are designated as SLOWPOKE-2 reactors. These research reactors are prone to two major issues: aging components and lack of relevance to a younger audience. In order to combat these problems, one SLOWPOKE -2 facility has embraced a strategy that involves modernizing their reactor in order to keep the reactor up to date and relevant. In 2001, this facility replaced its aging analogue reactor control system with a digital control system. The system was successfully commissioned and has provided a renewed platform for student learning and research. The digital control system provides a better interface and allows flexibility in data storage and retrieval that was never possible with the analogue control system. This facility has started work on another upgrade to the digital control and instrumentation system that will be installed in 2010. The upgrade includes new computer hardware, updated software and a web-based simulation and training system that will allow licensed operators, students and researchers to use an online simulation tool for training, education and research. The tool consists of: 1) A dynamic simulation for reactor kinetics (e.g., core flux, power, core temperatures, etc). This tool is useful for operator training and student education; 2) Dynamic mapping of the reactor and pool container gamma and neutron fluxes as well as the vertical neutron beam tube flux. This research planning tool is used for various researchers who wish to do irradiations (e.g., neutron

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)

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

Digital, remote control system for a 2-MW research reactor

International Nuclear Information System (INIS)

Battle, R.E.; Corbett, G.K.

1988-01-01

A fault-tolerant programmable logic controller (PLC) and operator workstations have been programmed to replace the hard-wired relay control system in the 2-MW Bulk Shielding Reactor (BSR) at Oak Ridge National Laboratory. In addition to the PLC and remote and local operator workstations, auxiliary systems for remote operation include a video system, an intercom system, and a fiber optic communication system. The remote control station, located at the High Flux Isotope Reactor 2.5 km from the BSR, has the capability of rector startup and power control. The system was designed with reliability and fail-safe features as important considerations. 4 refs., 3 figs

3D CAD model of the subcritical nuclear reactor of IPN

International Nuclear Information System (INIS)

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

2016-09-01

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

Techno-economic assessment of membrane assisted fluidized bed reactors for pure H_2 production with CO_2 capture

International Nuclear Information System (INIS)

Spallina, V.; Pandolfo, D.; Battistella, A.; Romano, M.C.; Van Sint Annaland, M.; Gallucci, F.

2016-01-01

Highlights: • Membrane reactors improve the overall efficiency of H_2 production up to 20%. • Respect to conventional reforming, the H_2 yield increases from 12% to 20%. • The COH is reduced of at least 220% using membrane reactors. • FBMR capture 72% of CO_2 with a specific cost of 8 eur/tonn_C_O_2_. • MA-CLR can reach 90% of CO_2 avoided with same cost of FTR. - Abstract: This paper addresses the techno-economic assessment of two membrane-based technologies for H_2 production from natural gas, fully integrated with CO_2 capture. In the first configuration, a fluidized bed membrane reactor (FBMR) is integrated in the H_2 plant: the natural gas reacts with steam in the catalytic bed and H_2 is simultaneously separated using Pd-based membranes, and the heat of reaction is provided to the system by feeding air as reactive sweep gas in part of the membranes and by burning part of the permeated H_2 (in order to avoid CO_2 emissions for heat supply). In the second system, named membrane assisted chemical looping reforming (MA-CLR), natural gas is converted in the fuel rector by reaction with steam and an oxygen carrier (chemical looping reforming), and the produced H_2 permeates through the membranes. The oxygen carrier is re-oxidized in a separate air reactor with air, which also provides the heat required for the endothermic reactions in the fuel reactor. The plants are optimized by varying the operating conditions of the reactors such as temperature, pressures (both at feed and permeate side), steam-to-carbon ratio and the heat recovery configuration. The plant design is carried out using Aspen Simulation, while the novel reactor concepts have been designed and their performance have been studied with a dedicated phenomenological model in Matlab. Both configurations have been designed and compared with reference technologies for H_2 production based on conventional fired tubular reforming (FTR) with and without CO_2 capture. The results of the analysis show

Comparative studies of ENDF/B-6.8, JEF-2.2 and JENDL-3.2 data libraries by monte carlo modeling of high temperature reactors on plutonium based fuel cycles

International Nuclear Information System (INIS)

Talamo, Alberto; Gudowski, Waclaw; Cetnar, Jerzy

2004-01-01

We performed a numerical comparative analysis of the burnup capability of the Gas Turbine-Modular Helium Reactor (GT-MHR) by the Monte Carlo Continuous Energy Burnup Code (MCB). The MCB code is an extension of MCNP that includes the burnup implementation; it adopts continuous energy cross sections and it evaluates the transmutation trajectories for over 2,400 decaying nuclides. We equipped the MCB code with three different nuclear data libraries: JENDL-3.2, JEF-2.2 and ENDF/B-6.8 processed for temperatures from 300 to 1,800K. The GT-MHR model studied in this paper is fueled by actinides coming from the Light Water Reactors waste, converted into two different types of fuel: Driver Fuel and Transmutation Fuel. The Driver Fuel supplies the fissile nuclides needed to maintain the criticality of the reactor, whereas the Transmutation Fuel depletes non-fissile isotopes and controls reactivity excess. We set the refueling and shuffling period to one year and the in-core fuel residency time to three years. The comparative analysis of the MCB code consists of accuracy and precision studies. In the accuracy studies, we performed the burnup calculation with different nuclear data libraries during the year at which the refueling and shuffling schedule set the equilibrium of the fuel composition. In the precision studies, we repeated the same simulations 20 times with a different pseudorandom number stride and the same nuclear data library. (author)

Nuclear reactor (1960)

International Nuclear Information System (INIS)

Maillard, M.L.

1960-01-01

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

Decontamination and Decommissioning Project of the TRIGA Mark - 2 and 3 research reactors

Energy Technology Data Exchange (ETDEWEB)

Jung, K. J.; Baik, S. T.; Chung, U. S.; Park, S. K.; Moon, J. S.; Jung, K. H.; Lee, B. J.; Kim, J. K.; Kim, K. H

1999-02-01

Design work for the D and D began in 1998, and expected to be finished at the end of February 1999. Base on the D and D design, the decommissioning plan and the environmental impact assessment report have been completed and submitted to the MOST for licensing at the end of 1998. These documents are being reviewing at the KINS. It is expected that the licensing from authority will be come out at the end of September 1999. Then D and D practical work will be started in the latter half of the year 1999 and completed by 2002. The first practical work of the D and D will be the KRR-2 reactor hall to transform the hall as a temporary storage of radioactive waste produced during the D and D work. Meanwhile, all the spent fuel from KRR-1 and 2 were safely transported to us at the middle of 1998. The D and D project was originally planned, to be finished by the end of the year 1999. This project were lately modified and extended until the end of the year 2002 because of the interim storage of radioactive wastes arising from the D and D work. These radioactive wastes will be stored at the KRR-2 reactor hall until a disposal facility is operational. (author)

Microflow photochemistry: UVC-induced [2 + 2]-photoadditions to furanone in a microcapillary reactor

Directory of Open Access Journals (Sweden)

Sylvestre Bachollet

2013-10-01

Full Text Available [2 + 2]-Cycloadditions of cyclopentene and 2,3-dimethylbut-2-ene to furanone were investigated under continuous-flow conditions. Irradiations were conducted in a FEP-microcapillary module which was placed in a Rayonet chamber photoreactor equipped with low wattage UVC-lamps. Conversion rates and isolated yields were compared to analogue batch reactions in a quartz test tube. In all cases examined, the microcapillary reactor furnished faster conversions and improved product qualities.

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

International Nuclear Information System (INIS)

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

2009-01-01

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

The BR2 materials testing reactor. Past, ongoing and under-study upgradings

Energy Technology Data Exchange (ETDEWEB)

Baugnet, J M; Roedt, Ch de; Gubel, P; Koonen, E [Centre d' Etude de I' Energie Nucleaire, Studiecentrum voor Kernenergie, C.E.N./S.C.K., Mol (Belgium)

1990-05-01

The BR2 reactor (Mol, Belgium) is a high-flux materials testing reactor. The fuel is 93% {sup 235}U enriched uranium. The nominal power ranges from 60 to 100 MW. The main features of the design are the following: 1) maximum neutron flux, thermal: 1.2 x 10{sup 15} n/cm{sup 2} s; fast (E > 0.1 MeV) : 8.4 x 10{sup 14} n /cm{sup 2} s; 2) great flexibility of utilization: the core configuration and operation mode can be adapted to the experimental loading; 3) neutron spectrum tailoring; 4) availability of five 200 mm diameter channels besides the standard channels (84 mm diameter); 5) access to the top and bottom covers of the reactor authorizing the irradiation of loops. The reactor is used to study the behaviour of fuel elements and structural materials intended for future nuclear power stations of several types (fission and fusion). Irradiations are carried out in connection with performance tests up to very high burn-up or neutron fluence as well as for safety experiments, power cycling experiments, and generally speaking, tests under off-normal conditions. Irradiations for nuclear transmutation (production of high specific activity radio-isotopes and transplutonium elements), neutron-radiography, use of beam tubes for physics studies, and gamma irradiations are also carried out. The BR2 is used in support of Belgian programs, at the request of utilities, industry and universities and in the framework of international agreements. The paper reviews the past and ongoing upgrading and enhancement of reactor capabilities as well as those under study or consideration, namely with regard to: reactor equipment, fuel elements, irradiation facilities, reactor operation conditions and long-term strategy. (author)

Reactor Physics Programme

Energy Technology Data Exchange (ETDEWEB)

De Raedt, C

2000-07-01

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

Reactor Physics Programme

International Nuclear Information System (INIS)

De Raedt, C.

2000-01-01

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

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)

Activation analysis of Al2O3 samples at the RB reactor

International Nuclear Information System (INIS)

Sokcic-Kostic, M.; Antic, D.; Pesic, M.

1994-01-01

Low flux activation analysis was successfully performed at the zero power RB reactor on a specific sample giving good results. For analysis of low concentration, isotopes must have cross sections of a few barns and periods of a few days with profitable disintegration schemes. This means irradiation of n x ∼ 10 min and flux n x 10 7 n/cm 2 s. The specific results of this and similar experiments provide information for industrial treatment of trace elements and sample analyses in different fields of research as well as food sample analysis

Thermodynamic analysis of behaviour of boiling water reactor coolant on the basis of solubility in Fe3O4-H2O-O2 system

International Nuclear Information System (INIS)

Zarembo, V.I.; Slobodov, A.A.; Kritskij, V.G.; Puchkov, L.V.; Sedov, V.M.

1986-01-01

The thermodynamic analysis of the behaviour of boiling water reactor coolant on the basis of solubility in Fe 3 O 4 -H 2 O-O 2 system is performed for the purpose of establishing the iron existence forms in non-sedimentated suspended corrosion product particles as well as iron concentration of corrosion origin in power plants. It is shown that the iron solubility in the considered system with temperature variation occurs through the maximum at 423 K. Below this temperature the crystal Fe(OH) 3 is responsible for its value, at higher temperatures - magnetite. The growth of equilibrium oxygen concentration from 0.1 to 1000 μg/kg H 2 O only slightly increases the magnetite solubility

The main pump motor remote visual check in the application of the domestic nuclear power plants

International Nuclear Information System (INIS)

Ge Lianwei; Yu Tao; Fang Jiang; Zhang Ting; Zhang Xingtian; Ding Youyuan

2014-01-01

In this paper, the Qinshan nuclear power station the first main pump motor to the successful implementation of remote visual inspection the main pump motor remote visual inspection applications. Qinshan Nuclear Power Plant Units 1 and 2 of the main pump motor inspection results show that the key components of the Qinshan Nuclear Power Plant Units 1 and 2 of the main pump rotor, stator end coils good condition, its problems for 10 years in the motor does not affect the normal use of the motor state disintegration overhaul problems tracking disintegration overhaul in 10 years. (authors)

UO{sub 2} and PuO{sub 2} utilization in high temperature engineering test reactor with helium coolant

Energy Technology Data Exchange (ETDEWEB)

Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Novitrian,; Pramuditya, Syeilendra; Su’ud, Zaki [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Aji, Indarta K. [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia)

2016-03-11

High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO{sub 2} fuel. In this study, we have evaluated the use of UO{sub 2} and PuO{sub 2} in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of {sup 235}U in loaded fuel is 18.0% or above.

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

Science.gov (United States)

Han, Jong-Hyun; Yoon, Do-Young

2010-06-01

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

Economics and utilization of thorium in nuclear reactors. Technical annexes 1 and 2

International Nuclear Information System (INIS)

1978-05-01

An assessment of the impact of utilizing the 233 U/thorium fuel cycle in the U.S. nuclear economy is strongly dependent upon several decisions involving nuclear energy policy. These decisions include: (1) to recycle or not recycle fissile material; (2) if fissile material is recycled, to recycle plutonium, 233 U, or both; and (3) to deploy or not to deploy advanced reactor designs such as Fast Breeder Reactors (FBR's), High Temperature Gas Reactors (HTGR's), and Canadian Deuterium Uranium Reactors (CANDU's). This report examines the role of thorium in the context of the above policy decisions while focusing special attention on economics and resource utilization

TPDWR2: thermal power determination for Westinghouse reactors, Version 2. User's guide

International Nuclear Information System (INIS)

Kaczynski, G.M.; Woodruff, R.W.

1985-12-01

TPDWR2 is a computer program which was developed to determine the amount of thermal power generated by any Westinghouse nuclear power plant. From system conditions, TPDWR2 calculates enthalpies of water and steam and the power transferred to or from various components in the reactor coolant system and to or from the chemical and volume control system. From these results and assuming that the reactor core is operating at constant power and is at thermal equilibrium, TPDWR2 calculates the thermal power generated by the reactor core. TPDWR2 runs on the IBM PC and XT computers when IBM Personal Computer DOS, Version 2.00 or 2.10, and IBM Personal Computer Basic, Version D2.00 or D2.10, are stored on the same diskette with TPDWR2

Once-through CANDU reactor models for the ORIGEN2 computer code

International Nuclear Information System (INIS)

Croff, A.G.; Bjerke, M.A.

1980-11-01

Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % 235 U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given

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

International Nuclear Information System (INIS)

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

1974-01-01

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

Feasibility Study for Cobalt Bundle Loading to CANDU Reactor Core

International Nuclear Information System (INIS)

Park, Donghwan; Kim, Youngae; Kim, Sungmin

2016-01-01

CANDU units are generally used to produce cobalt-60 at Bruce and Point Lepreau in Canada and Embalse in Argentina. China has started production of cobalt-60 using its CANDU 6 Qinshan Phase III nuclear power plant in 2009. For cobalt-60 production, the reactor’s full complement of stainless steel adjusters is replaced with neutronically equivalent cobalt-59 adjusters, which are essentially invisible to reactor operation. With its very high neutron flux and optimized fuel burn-up, the CANDU has a very high cobalt-60 production rate in a relatively short time. This makes CANDU an excellent vehicle for bulk cobalt-60 production. Several studies have been performed to produce cobalt-60 using adjuster rod at Wolsong nuclear power plant. This study proposed new concept for producing cobalt-60 and performed the feasibility study. Bundle typed cobalt loading concept is proposed and evaluated the feasibility to fuel management without physics and system design change. The requirement to load cobalt bundle to the core was considered and several channels are nominated. The production of cobalt-60 source is very depend on the flux level and burnup directly. But the neutron absorption characteristic of cobalt bundle is too high, so optimizing design study is needed in the future

Feasibility Study for Cobalt Bundle Loading to CANDU Reactor Core

Energy Technology Data Exchange (ETDEWEB)

Park, Donghwan; Kim, Youngae; Kim, Sungmin [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2016-10-15

CANDU units are generally used to produce cobalt-60 at Bruce and Point Lepreau in Canada and Embalse in Argentina. China has started production of cobalt-60 using its CANDU 6 Qinshan Phase III nuclear power plant in 2009. For cobalt-60 production, the reactor’s full complement of stainless steel adjusters is replaced with neutronically equivalent cobalt-59 adjusters, which are essentially invisible to reactor operation. With its very high neutron flux and optimized fuel burn-up, the CANDU has a very high cobalt-60 production rate in a relatively short time. This makes CANDU an excellent vehicle for bulk cobalt-60 production. Several studies have been performed to produce cobalt-60 using adjuster rod at Wolsong nuclear power plant. This study proposed new concept for producing cobalt-60 and performed the feasibility study. Bundle typed cobalt loading concept is proposed and evaluated the feasibility to fuel management without physics and system design change. The requirement to load cobalt bundle to the core was considered and several channels are nominated. The production of cobalt-60 source is very depend on the flux level and burnup directly. But the neutron absorption characteristic of cobalt bundle is too high, so optimizing design study is needed in the future.

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

Thermal design of heat-exchangeable reactors using a dry-sorbent CO2 capture multi-step process

International Nuclear Information System (INIS)

Moon, Hokyu; Yoo, Hoanju; Seo, Hwimin; Park, Yong-Ki; Cho, Hyung Hee

2015-01-01

The present study proposes a multi-stage CO 2 capture process that incorporates heat-exchangeable fluidized-bed reactors. For continuous multi-stage heat exchange, three dry regenerable sorbents: K 2 CO 3 , MgO, and CaO, were used to create a three-stage temperature-dependent reaction chain for CO 2 capture, corresponding to low (50–150 °C), middle (350–650 °C), and high (750–900 °C) temperature stages, respectively. Heat from carbonation in the high and middle temperature stages was used for regeneration for the middle and low temperature stages. The feasibility of this process is depending on the heat-transfer performance of the heat-exchangeable fluidized bed reactors as the focus of this study. The three-stage CO 2 capture process for a 60 Nm 3 /h CO 2 flow rate required a reactor area of 0.129 and 0.130 m 2 for heat exchange between the mid-temperature carbonation and low-temperature regeneration stages and between the high-temperature carbonation and mid-temperature regeneration stages, respectively. The reactor diameter was selected to provide dense fluidization conditions for each bed with respect to the desired flow rate. The flow characteristics and energy balance of the reactors were confirmed using computational fluid dynamics and thermodynamic analysis, respectively. - Highlights: • CO 2 capture process is proposed using a multi-stage process. • Reactor design is conducted considering heat exchangeable scheme. • Reactor surface is designed by heat transfer characteristics of fluidized bed

Real-Time Quantitative Operando Raman Spectroscopy of a CrOx/Al2O3 Propane Dehydrogenation Catalyst in a Pilot-Scale Reactor

NARCIS (Netherlands)

Sattler, Jesper J. H. B.; Mens, Ad M.; Weckhuysen, Bert M.

2014-01-01

Combined operando UV/vis-Raman spectroscopy has been used to study the deactivation of CrOx/Al2O3 catalyst extrudates in a pilot scale propane dehydrogenation reactor. For this purpose, UV/vis and Raman optical fiber probes have been designed, constructed and tested. The light absorption measured by

Modelling of MOCVD Reactor: New 3D Approach

Science.gov (United States)

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

2014-04-01

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

Modelling of MOCVD reactor: new 3D approach

International Nuclear Information System (INIS)

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

2014-01-01

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

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

Science.gov (United States)

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

1993-11-01

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

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)

Safe dismantling of the SVAFO research reactors R2 and R2-0 in Sweden

International Nuclear Information System (INIS)

ARNOLD, Hans-Uwe; BROY, Yvonne; Dirk Schneider

2017-01-01

The R2 and R2-0 reactors were part of the Swedish government's research program on nuclear power from the early 1960's. Both reactors were shut down in 2005 following a decision by former operator Studsvik Nuclear AB. The decommissioning of the R2 and R2-0 reactors is divided into three phases. The first phase - awarded to AREVA - involved dismantling of the reactors and associated systems in the reactor pool, treatment of the disassembled components as well as draining, cleaning and emptying the pool. In the second phase, the pool structure itself will be dismantled, while removal of remaining reactor systems, treatment and disposal of materials and clean-up will be carried out in the third stage. The entire work is planned to be completed before the end of this decade. The paper describes the several steps of phase 1 - starting with the team building, followed by the dismantling operations and covers challenges encountered and lessons learned as well. The reactors consist of 5.400 kg aluminum, 6.000 kg stainless steel restraint structures as well as, connection elements of the mostly flanged components (1.000 kg). The most demanding - from a radiological point of view - was the R2-0 reactor that was limited to ∼ 1 m"3 construction volumes but with an extremely heterogeneous activation profile. Based on the calculated radiological entrance data and later sampling, nuclide vectors for both reactors depending on the real placement of the single component and on the material (aluminum and stainless steel) were created. Finally, for the highest activated component from R2 reactor, 85 Sv/h were measured. The dismantling principles - adopted on a safety point of view - were the following: The always protected base area of the ponds served as a flexible buffer area for waste components and packaging. Specific protections were also installed on the walls to protect them from mechanical stress which may occur during dismantling work. A specific work platform was

IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-07-01

The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1.

IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

International Nuclear Information System (INIS)

1992-01-01

The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1

HERESY, 2-D Few-Group Static Eigenvalues Calculation for Thermal Reactor

International Nuclear Information System (INIS)

Finch, D.R.

1965-01-01

1 - Description of problem or function: HERESY3 solves the two- dimensional, few-group, static reactor eigenvalue problem using the heterogeneous (source-sink or Feinburg-Galanin) formalism. The solution yields the reactor k-effective and absorption reaction rates for each rod normalized to the most absorptive rod in the thermal level. Epithermal fissions are allowed at each resonance level, and lattice-averaged values of thermal utilization, resonance escape probability, thermal and resonance eta values, and the fast fission factor are calculated. Kernels in the calculation are based on age-diffusion theory. Both finite reactor lattices and infinitely repeating reactor super-cells may be calculated. Rod parameters may be calculated by several internal options, and a direct interface is provided to a HAMMER system (NESC Abstract 277) lattice library tape to obtain cell parameters. Criticality searches are provided on thermal utilization, thermal eta, and axial leakage buckling. 2 - Method of solution: Direct power iteration on matrix form of the heterogeneous critical equation is used. 3 - Restrictions on the complexity of the problem: Maxima of - 50 flux/geometry symmetry positions; 20 physically different assemblies; 9 resonance levels; 5000 rod coordinate positions

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

International Nuclear Information System (INIS)

1983-04-01

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

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

International Nuclear Information System (INIS)

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

1993-01-01

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

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

International Nuclear Information System (INIS)

Chiba, Gou

2002-12-01

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

Status and perspective of development of cold moderators at the IBR-2 reactor

International Nuclear Information System (INIS)

Kulikov, S; Shabalin, E

2012-01-01

The modernized IBR-2M reactor will start its operation with three water grooved moderators in 2011. Afterwards, they will be exchanged by a new complex of moderators. The complex consists of three so-called kombi-moderators, each of them containing a pre-moderator, a cold moderator, grooved ambient water moderators and post-moderators. They are mounted onto three moveable trolleys that serve to deliver and install moderators near the reactor core. The project is divided in three stages. In 2012 the first stage of development of complex of moderators will be finished. The water grooved moderator will be replaced with the new kombi-moderator for beams nos. 7, 8, 10, 11. Main parameters of moderators for this direction will be studied then. The next stages will be done for beams nos. 2-3 and for beams nos. 1, 9, 4-6, consequently. Cold moderator chambers at the modernized IBR-2 reactor are filled with thousands of beads (∼3.5 - 4 mm in diameter) of moderating material. The cold helium gas flow delivers beads from the charging device to the moderator during the fulfillment process and cools down them during the reactor cycle. The mixture of aromatic hydrocarbons (mesithylen and m-xylen) has been chosen as moderating material. The explanation of the choice of material for novel cold neutron moderators, configuration of moderator complex for the modernized IBR-2 reactor and the main results of optimization of moderator complex for the third stage of moderator development are discussed in the article.

Status and perspective of development of cold moderators at the IBR-2 reactor

Science.gov (United States)

Kulikov, S.; Shabalin, E.

2012-03-01

The modernized IBR-2M reactor will start its operation with three water grooved moderators in 2011. Afterwards, they will be exchanged by a new complex of moderators. The complex consists of three so-called kombi-moderators, each of them containing a pre-moderator, a cold moderator, grooved ambient water moderators and post-moderators. They are mounted onto three moveable trolleys that serve to deliver and install moderators near the reactor core. The project is divided in three stages. In 2012 the first stage of development of complex of moderators will be finished. The water grooved moderator will be replaced with the new kombi-moderator for beams #7, 8, 10, 11. Main parameters of moderators for this direction will be studied then. The next stages will be done for beams #2-3 and for beams #1, 9, 4-6, consequently. Cold moderator chambers at the modernized IBR-2 reactor are filled with thousands of beads (~3.5 - 4 mm in diameter) of moderating material. The cold helium gas flow delivers beads from the charging device to the moderator during the fulfillment process and cools down them during the reactor cycle. The mixture of aromatic hydrocarbons (mesithylen and m-xylen) has been chosen as moderating material. The explanation of the choice of material for novel cold neutron moderators, configuration of moderator complex for the modernized IBR-2 reactor and the main results of optimization of moderator complex for the third stage of moderator development are discussed in the article.

Mass transfer of ammonia escape and CO2 absorption in CO2 capture using ammonia solution in bubbling reactor

International Nuclear Information System (INIS)

Ma, Shuangchen; Chen, Gongda; Zhu, Sijie; Han, Tingting; Yu, Weijing

2016-01-01

Highlights: • Mass transfer coefficient models of ammonia escape were built. • Influences of temperature, inlet CO 2 and ammonia concentration were studied. • Mass transfer coefficients of ammonia escape and CO 2 absorption were obtained. • Studies can provide the basic data as a reference guideline for process application. - Abstract: The mass transfer of CO 2 capture using ammonia solution in the bubbling reactor was studied; according to double film theory, the mass transfer coefficient models and interface area model were built. Through our experiments, the overall volumetric mass transfer coefficients were obtained, while the interface areas in unit volume were estimated. The volumetric mass transfer coefficients of ammonia escaping during the experiment were 1.39 × 10 −5 –4.34 × 10 −5 mol/(m 3 s Pa), and the volumetric mass transfer coefficients of CO 2 absorption were 2.86 × 10 −5 –17.9 × 10 −5 mol/(m 3 s Pa). The estimated interface area of unit volume in the bubbling reactor ranged from 75.19 to 256.41 m 2 /m 3 , making the bubbling reactor a viable choice to obtain higher mass transfer performance than the packed tower or spraying tower.

Characterization of the Three Mile Island Unit-2 reactor building atmosphere prior to the reactor building purge

International Nuclear Information System (INIS)

Hartwell, J.K.; Mandler, J.W.; Duce, S.W.; Motes, B.G.

1981-05-01

The Three Mile Island Unit-2 reactor building atmosphere was sampled prior to the reactor building purge. Samples of the containment atmosphere were obtained using specialized sampling equipment installed through penetration R-626 at the 358-foot (109-meter) level of the TMI-2 reactor building. The samples were subsequently analyzed for radionuclide concentration and for gaseous molecular components (O 2 , N 2 , etc.) by two independent laboratories at the Idaho National Engineering Laboratory (INEL). The sampling procedures, analysis methods, and results are summarized

Effect of kinetic parameters on simultaneous ramp reactivity insertion plus beam tube flooding accident in a typical low enriched U{sub 3}Si{sub 2}-Al fuel-based material testing reactor-type research reactor

Energy Technology Data Exchange (ETDEWEB)

Nasir, Rubina; Mirza, Nasir M. [Dept. of, Physics, Air University, Islamabad (Pakistan); Mirza, Sikander M. [Dept. of, Physics and Applied Mathematics, Pakistan Institute of Engineering and Applied Sciences, Post Office Nilore, Islamabad (Pakistan)

2017-06-15

This work looks at the effect of changes in kinetic parameters on simultaneous reactivity insertions and beam tube flooding in a typical material testing reactor-type research reactor with low enriched high density (U{sub 3}Si{sub 2}-Al) fuel. Using a modified PARET code, various ramp reactivity insertions (from $0.1/0.5 s to $1.3/0.5 s) plus beam tube flooding ($0.5/0.25 s) accidents under uncontrolled conditions were analyzed to find their effects on peak power, net reactivity, and temperature. Then, the effects of changes in kinetic parameters including the Doppler coefficient, prompt neutron lifetime, and delayed neutron fractions on simultaneous reactivity insertion and beam tube flooding accidents were analyzed. Results show that the power peak values are significantly sensitive to the Doppler coefficient of the system in coupled accidents. The material testing reactor-type system under such a coupled accident is not very sensitive to changes in the prompt neutron life time; the core under such a coupled transient is not very sensitive to changes in the effective delayed neutron fraction.

First results of U3Si2 production and its relevance in the power scale-up of IPEN research reactor IEA-R1m

International Nuclear Information System (INIS)

Saliba-Silva, A.M.; Souza, J.A.B.; Frajndlich, E.U.C.; Durazzo, M.; Perrotta, J.A.

1997-01-01

The own supply of LEU U 3 Si 2 is crucial for IPEN, since the whole scale-up of IPEN MTR IEA-Rlm reactor will rely on it. The Brazilian request for radioisotopes production is fully linked with the already made power scale-up from 2 to 5 MW for this reactor. IPEN now depends on fuel element material upgrading from U 3 O 8 towards LEU U 3 Si 2 . The fuel plate productive technology from the powdered material is already well established, only needing simple making of minor adjustments, but to reach the stage of producing U 3 Si 2 we need a fully settled chemical pilot plant in order to reach a LEU UF 4 productive routine. Complementing this process, it was also needed to scale down the previous practice of uranium magnesiothermic reduction to around a sub-critical safe uranium mass of approximately 3000g. To complete the metallurgical processing, it is being developed the production of U 3 Si 2 in a vacuum induction furnace. Some experiments to get this intermetallic, using natural uranium, have already been carried out in order to build up a general idea of the future process of LEU U 3 Si 2 . These experiments are described in this paper and also some of the initial characterization results, such as the qualification pattern of the ingot. It is also discussed some new features of inhomogeneity of solidified phases that may be deleterious to future production routine. (author)

Analysis of SBO accident and natural circulation of 49-2 swimming pool reactor

International Nuclear Information System (INIS)

Wu Yuanyuan; Liu Tiancai; Sun Wei

2012-01-01

The transient thermal hydraulic characteristics of 49-2 Swimming Pool Reactor (SPR) were analyzed by RELAP5/MOD3.3 code to verify the capability of natural circulation and minus reactivity feedback for accident mitigation under the condition of station blackout (SBO). Then, the effects on accident consequence and sequence for core channels and primary pumps were briefly discussed. The calculation results show that the reactor can be shutdown by the effect of minus reactivity feedback, and the residual heat can be removed through the stable natural circulation. Therefore, it demonstrates that the 49-2 SPR is safe during the accident of SBO. (authors)

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1969-02-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1969-02-01

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

YBa2Cu3O7-δ thin films deposited by MOCVD vertical reactor with a flow guide

International Nuclear Information System (INIS)

Sujiono, E.H.; Negeri Makassar; Sani, R.A.; Saragi, T.; Arifin, P.; Barmawi, M.

2001-01-01

The effect of a flow guide in a vertical MOCVD reactor on the deposition uniformity and growth rate of thin YBCO films has been studied. Without the flow guide the growth rates are low, have a poor uniformity and the film composition is not stoichiometric. The growth rate of the films grown using a reactor with the flow guide was approximately twice that without the flow guide. Using this flow guide the growth rates were 0.4-0.7 μm/h for growth temperatures varying between 600 and 750 C, and the crystalline quality as well as the surface morphology of YBCO films on MgO substrates is improved. For films grown at temperatures above 650 C the composition of Y:Ba:Cu is 1:2:3, as confirmed by EDAX spectra. Films deposited without and with the flow guide at 700 C have critical temperatures around 85 and 88 K, respectively. The reduction in ΔT c (T c,zero -T c,onset ) also shows an improvement of the superconducting properties of YBCO thin films deposited with a flow guide. (orig.)

G 2 reactor project; Projet de pile a double fin: G 2

Energy Technology Data Exchange (ETDEWEB)

Ailleret, [Electricite de France (EDF), Dir. General des Etudes de Recherches, 75 - Paris (France); Taranger, P; Yvon, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

The CEA actually constructs the G-2 reactor core working with natural uranium, which will use graphite as moderator, and gas under pressure as cooling fluid. This report presents the specificity of the new reactor: - the different elements of the reactor core, - the control and the security of the reactor, - the renewal of the fuel, - the biologic surrounding wall, - and the cooling circuit. (M.B.) [French] le Commissariat a l'Energie Atomique construit actuellement la pile G-2 a Uranium naturel, qui utilisera le graphite comme moderateur, et le gaz sous pression comme fluide de refroidissement. Ce rapport presente les specificite du nouveau reacteur: - les differents elements de la pile, - le controle et la securite du reacteur, - le renouvellement du combustible, - l'enceinte biologique, - et le circuit de refroidissement. (M.B.)

Reactor handbook. 2. rev. ed.

International Nuclear Information System (INIS)

Lederer, B.J.; Wildberg, D.W.

1992-01-01

On the basis of the guidelines on expert knowledge, the book discusses the subjects of atomic physics, heat transfer, nuclear power plants, reactor materials, radiation protection, reactor safety, reactor instrumentation, and reactor operation, with special regard to nuclear power plants with LWR-type reactors. The book is intended for shift personnel, especially gang bosses, reactor operators, and control station operators: for this reason a practical and rather popular style has been chosen. However, the book will also be a manual for other operating personnel, personnel of producer companies, expert organisations, authorities, and students. It can be used as a textbook for staff training, a manual for the practice, and as accompanying book for teaching at nuclear engineering schools. (orig.) With 173 figs [de

Descriptive data of JENDL-3.2

Energy Technology Data Exchange (ETDEWEB)

Shibata, Keiichi; Narita, Tsutomu [eds.; Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-02-01

The second revision of JENDL-3 (JENDL-3.2) was released in 1994. The library contains evaluated neutron nuclear data for 340 nuclides which are needed for the design of fission and fusion reactors and for shielding calculation. This report presents a brief description of the evaluation method which is given in the File-1 part of JENDL-3.2. (author)

Annual report on JEN-1 and JEN-2 Reactors; Informe periodico de Reactores JEN-1 y JEN-2 correpondiente al ano 1972

Energy Technology Data Exchange (ETDEWEB)

Montes Ponce de Leon, J.

1974-07-01

In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

Manpower development for safe operation of nuclear power plant. China. Simulator training for instructions. Activity: 2.1.4-Task-16. Technical report

International Nuclear Information System (INIS)

Han, Dong Hyun; Song, Suk Ill.

1996-01-01

By the request of the Qinshan Nuclear Power training center, Korea Electric Power Company (KEPCO) expert team visited the Qinshan Nuclear Power Training Center during October 7-21, 1996. The purpose of the visiting was as follows: To give some ideas, through KEPCO KNTC training experiences about operator training programme including simulator training - how to improve simulator instructors' training skill and knowledge; how to conduct classroom and simulator lectures; how to prepare lesson note for lectures; how to make the trainees evaluation; how to course analyze and feed back; how to make scenario for simulator training. To fulfill above purposes, the expert team used KNTC procedures, 1996 KNTC training plan, development and qualification for instructor, simulator training and evaluation, control and preparedness of lesson notes. These procedures were used only to establish the framework for Qinshan nuclear training center's procedures

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

International Nuclear Information System (INIS)

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

1993-11-01

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

Analysis and relevant treatment of diametral tolerance of exciter shaft in unit 2

International Nuclear Information System (INIS)

Liu Qiang

2012-01-01

The generator and exciter unit has three support in Qinshan Nuclear Power Plant Phase Ⅱ, there are two bearings for the generator rotor and one for the exciter, this structure results in that it is difficult to achieve the standard when checking the exciter bearing's diametral tolerance. In the fifth outage of unit 2 in Qinshan Nuclear Power Plant Phase Ⅱ, the diametral tolerance failed to achieve the standard, there were several reasons for this, such as the alignment of generator and exciter coupling, the angular moment of generator and exciter coupling bolt. the end surface condition of generator and exciter coupling, the fitting dimension of the coupling bolt hole and the sleeve in it. After the analysis of all reasons one by one, it was confirmed that the radical reason was the abnormal condition of the generator coupling end surface, the problem was solved by machining it. (author)

Research on Primary Shielding Calculation Source Generation Codes

Science.gov (United States)

Zheng, Zheng; Mei, Qiliang; Li, Hui; Shangguan, Danhua; Zhang, Guangchun

2017-09-01

Primary Shielding Calculation (PSC) plays an important role in reactor shielding design and analysis. In order to facilitate PSC, a source generation code is developed to generate cumulative distribution functions (CDF) for the source particle sample code of the J Monte Carlo Transport (JMCT) code, and a source particle sample code is deveoped to sample source particle directions, types, coordinates, energy and weights from the CDFs. A source generation code is developed to transform three dimensional (3D) power distributions in xyz geometry to source distributions in r θ z geometry for the J Discrete Ordinate Transport (JSNT) code. Validation on PSC model of Qinshan No.1 nuclear power plant (NPP), CAP1400 and CAP1700 reactors are performed. Numerical results show that the theoretical model and the codes are both correct.

Advanced I and C system of security level for nuclear power station

International Nuclear Information System (INIS)

Liu Yanyang

2001-01-01

Advanced I and C system of security level using for PWR developed by Framatome and Schneider collective, SPINLINE3, are introduced. The technology is used to outside reactor nuclear measurement system in Qinshan II period. It's succeed benefits by Framatome and Schneider's more years development experience in nuclear power station digitallization security level I and C system field, which improve security and reliability of PWR, and, easy operation and maintains. SPINLINE3 based on digitallization and modularization technical proposal, and covered entireness reactor protect system and correlative control system. The paper also introduce CLARISSE (computer aided design aid) and SCADE (embedded software aid) for developing SPINLINE3. SPINLINE3 fills correlative IS and rule, based on software and hardware unit which certificate and launch into operation. After brief review of Framatome and Schneider's experience, the paper are introducing design guideline, application technology and how to fill demand of security level I and C system

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.)

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Set of rules SOR 2 licensing of nuclear reactors

International Nuclear Information System (INIS)

1976-05-01

This is the set of rules promulgated by the Israel Atomic Energy Commission pursuant to the Supervision of Supplies and Services Law 5718-1957, Order regarding Supervision of Nuclear Reactors (1974) Chapter 3: Permits, to provide for the Licensing of Nuclear Reactors. (B.G.)

The restart of Belgium reactors of Doel 3 and Tihange 2. Doel 3 and Tihange 2: indications of defects in vessel steel

International Nuclear Information System (INIS)

2015-01-01

In a first part, an IRSN report comments the issue of restarting some Belgium reactors after the detection of defects (due to the presence of hydrogen) in the vessel steel of reactors during the third decennial inspection by Electrabel. The report describes the procedure followed by Electrabel and the Belgium nuclear authority (AFCN) to confirm that the detected defects were not harmful. It comments the defect detection and characterisation, the origin and potential evolution of defects, the assessment of mechanical characteristics of some components, the assessment of the defect harmfulness in terms of failure risk, and additional measurements. The second part contains the AFCN report which addresses: the chronology and scientific context, the actors, the situation of other Belgium reactors, an indication of published reports and press releases

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Upgrading of the research reactors FRG-1 and FRG-2

International Nuclear Information System (INIS)

Krull, W.

1981-01-01

In 1972 for the research reactor FRG-2 we applied for a license to increase the power from 15 MW to 21 MW. During this procedure a public laying out of the safety report and an upgrading procedure for both research reactors - FRG-1 (5 MW) and FRG-2 - were required by the licensing authorities. After discussing the legal background for licensing procedures in the Federal Republic of Germany the upgrading for both research reactors is described. The present status and future licensing aspects for changes of our research reactors are discussed, too. (orig.) [de

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

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-07-01

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

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

International Nuclear Information System (INIS)

Chang, Y.I.

1992-01-01

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

Performance analysis of photocatalytic CO2 reduction in optical fiber monolith reactor with multiple inverse lights

International Nuclear Information System (INIS)

Yuan, Kai; Yang, Lijun; Du, Xiaoze; Yang, Yongping

2014-01-01

Highlights: • A new optical fiber monolith reactor model for CO 2 reduction was developed. • Methanol concentration versus fiber location and operation parameters was obtained. • Reaction efficiency increases by 31.1% due to the four fibers and inverse layout. • With increasing space of fiber and channel center, methanol concentration increases. • Methanol concentration increases as the vapor ratio and light intensity increase. - Abstract: Photocatalytic CO 2 reduction seems potential to mitigate greenhouse gas emissions and produce renewable energy. A new model of photocatalytic CO 2 reduction in optical fiber monolith reactor with multiple inverse lights was developed in this study to improve the conversion of CO 2 to CH 3 OH. The new light distribution equation was derived, by which the light distribution was modeled and analyzed. The variations of CH 3 OH concentration with the fiber location and operation parameters were obtained by means of numerical simulation. The results show that the outlet CH 3 OH concentration is 31.1% higher than the previous model, which is attributed to the four fibers and inverse layout. With the increase of the distance between the fiber and the monolith center, the average CH 3 OH concentration increases. The average CH 3 OH concentration also rises as the light input and water vapor percentage increase, but declines with increasing the inlet velocity. The maximum conversion rate and quantum efficiency in the model are 0.235 μmol g −1 h −1 and 0.0177% respectively, both higher than previous internally illuminated monolith reactor (0.16 μmol g −1 h −1 and 0.012%). The optical fiber monolith reactor layout with multiple inverse lights is recommended in the design of photocatalytic reactor of CO 2 reduction

UCLA program in reactor studies: The ARIES tokamak reactor study

International Nuclear Information System (INIS)

1991-01-01

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

Abatement of fluorinated compounds using a 2.45 GHz microwave plasma torch with a reverse vortex plasma reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, J.H.; Cho, C.H.; Shin, D.H. [Plasma Technology Research Center, National Fusion Research Institute, 814-2 Oxikdo-dong, Gunsan-city, Jeollabuk-do (Korea, Republic of); Hong, Y.C., E-mail: ychong@nfri.re.kr [Plasma Technology Research Center, National Fusion Research Institute, 814-2 Oxikdo-dong, Gunsan-city, Jeollabuk-do (Korea, Republic of); Shin, Y.W. [Plasma Technology Research Center, National Fusion Research Institute, 814-2 Oxikdo-dong, Gunsan-city, Jeollabuk-do (Korea, Republic of); School of Advanced Green Energy and Environments, Handong Global University, Heunghae-eup, Buk-gu, Pohang-city, Gyeongbuk (Korea, Republic of)

2015-08-30

Highlights: • We developed a microwave plasma torch with reverse vortex reactor (RVR). • We calculated a volume fraction and temperature distribution of discharge gas and waste. • The performance of reverse vortex reactor increased from 29% to 43% than conventional vortex reactor. - Abstract: Abatement of fluorinated compounds (FCs) used in semiconductor and display industries has received an attention due to the increasingly stricter regulation on their emission. We have developed a 2.45 GHz microwave plasma torch with reverse vortex reactor (RVR). In order to design a reverse vortex plasma reactor, we calculated a volume fraction and temperature distribution of discharge gas and waste gas in RVR by ANSYS CFX of computational fluid dynamics (CFD) simulation code. Abatement experiments have been performed with respect to SF{sub 6}, NF{sub 3} by varying plasma power and N{sub 2} flow rates, and FCs concentration. Detailed experiments were conducted on the abatement of NF{sub 3} and SF{sub 6} in terms of destruction and removal efficiency (DRE) using Fourier transform infrared (FTIR). The DRE of 99.9% for NF{sub 3} was achieved without an additive gas at the N{sub 2} flow rate of 150 liter per minute (L/min) by applying a microwave power of 6 kW with RVR. Also, a DRE of SF{sub 6} was 99.99% at the N{sub 2} flow rate of 60 L/min using an applied microwave power of 6 kW. The performance of reverse vortex reactor increased about 43% of NF{sub 3} and 29% of SF{sub 6} abatements results definition by decomposition energy per liter more than conventional vortex reactor.

Investigation of tritium and 233U breeding in a fission-fusion hybrid reactor fuelling with ThO2

International Nuclear Information System (INIS)

Yildiz, K.; Sahin, S.; Sahin, H. M.; Acir, A.; Yalcin, S.; Altinok, T.; Bayrak, M.; Alkan, M.; Durukan, O.

2007-01-01

In the world, thorium reserves are three times more than natural Uranium reserves. It is planned in the near future that nuclear reactors will use thorium as a fuel. Thorium is not a fissile isotope because it doesn't make fission with thermal neutrons so it could be converted to 2 33U isotope which has very high quality fission cross-section with thermal neutrons. 2 33U isotope can be used in present LWRs as an enrichment fuel. In the fusion reactors, tritium is the most important fossil fuel. Because tritium is not natural isotope, it has to be produced in the reactor. The purpose of this work is to investigate the tritium and 2 33U breeding in a fission-fusion hybrid reactor fuelling with ThO 2 for Δt=10 days during a reactor operation period in five years. The neutronic analysis is performed on an experimental hybrid blanket geometry. In the center of the hybrid blanket, there is a line neutron source in a cylindrical cavity, which simulates the fusion plasma chamber where high energy neutrons (14.1 MeV) are produced. The conventional fusion reaction delivers the external neutron source for blankets following, 2 D + 3 T →? 4 He (3.5 MeV) + n (14.1 MeV). (1) The fuel zone made up of natural-ThO 2 fuel and it is cooled with different coolants. In this work, five different moderator materials, which are Li 2 BeF 4 , LiF-NaF-BeF 2 , Li 2 0Sn 8 0, natural Lithium and Li 1 7Pb 8 3, are used as coolants. The radial reflector, called tritium breeding zones, is made of different Lithium compounds and graphite in sandwich structure. In the work, eight different Lithium compounds were used as tritium breeders in the tritium breeding zones, which are Li 3 N, Li 2 O, Li 2 O 2 , Li 2 TiO 3 , Li 4 SiO 3 , Li 2 ZrO 3 , LiBr and LiH. Neutron transport calculations are conducted in spherical geometry with the help of SCALE4.4A SYSTEM by solving the Boltzmann transport equation with code CSAS and XSDRNPM, under consideration of unresolved and resolved resonances, in S 8 -P 3

Examination policy concerning the additional installation of No. 3 and No. 4 reactors in Takahama Nuclear Power Station and No. 3 and No. 4 reactors in Fukushima No. 2 Nuclear Power Station

International Nuclear Information System (INIS)

1980-01-01

The Nuclear Safety Commission decided the annual examination policy on the modification of reactor installation in Takahama Nuclear Power Station to construct No. 3 and No. 4 reactors inquired under date of November 26, 1979, by the Minister of International Trade and Industry, so that the examination results of the accident in Three Mile Island nuclear power station are reflected to the examination for the purpose of improving reactor safety. The examination results of the accident in Three Mile Island power station are being investigated by the Committee on Examination of Reactor Safety, based on the policy shown in ''On the second report of the special committee examining the accident in a nuclear power station in the U.S.'' determined by the Nuclear Safety Commission under date of September 13, 1979. Though the Committee will further clarify the past guideline about the items concerning the criteria, design and operation management, the Committee decided the tentative policy to reflect it to safety examination. Further, a table is attached, in which 52 items to be reflected to the security measures are classified from the viewpoint of necessity to reflect them to the final examination. This table includes 13 items of criteria and examination, 7 items related to design, 10 items related to operation management, 10 antidisaster items, and 12 items related to safety research. (Wakatsuki, Y.)

Comparison of ASTEC 1.3 and ASTEC 1.3 R2 calculations in case of SBO for VVER-1000 reactor

International Nuclear Information System (INIS)

Atanasova, B.; Stefanova, A.; Grudev, P.

2009-01-01

The report presents the results from severe accident analyses performed with the both versions of ASTEC v1.3 and ASTEC v1.3R2 computer code for a VVER 1000 type of reactor. The purpose of this analysis is to assess the progress of ASTEC code modeling of main phenomena arising during hypothetical severe accidents. The final target of these analyses is to estimate the behaviour of the ASTEC code, its capability for simulation of severe accidents, including safety systems and Severe Accident Management (SAM) procedures. The analyses have been performed assuming a station blackout with simultaneous loss of HPIS, LPIS (ECCSs), EFWS and spray system due to failure of DGs. Hydro accumulators are not available. In the calculation it is assumed opening and stuck-open of PRZ relief valves. It has been organized the Fission Products path through the SEMPELL valve. It should be said that this investigation was limited to the 'in-vessel' phase of the sequence; therefore the effect of sprays on containment atmosphere has not been studied. (authors)

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

Energy Technology Data Exchange (ETDEWEB)

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

2004-07-01

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

Annual report on JEN-1 and JEN-2 Reactors

International Nuclear Information System (INIS)

Montes Ponce de Leon, J.

1974-01-01

In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

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

International Nuclear Information System (INIS)

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

2003-01-01

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

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

International Nuclear Information System (INIS)

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

1996-06-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-06-01

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

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

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

Analysis of the metallic containment integrity of Angra 2/3 reactor under the effects of the design basis accident

International Nuclear Information System (INIS)

Costa, J.R.

1981-06-01

The application of Condru 4 computer code, developed to determine the maximum values of pressure and temperature that occur inside the metallic containment building of PWR nuclear power plants, in case of a hypothetic accident - LOCA - considered as a Design Basic Accident - DBA. The hypothesis, input and results for the simulation of a loss of coolant in the hot leg of the Angra-2/3 reactors, considered as the most critical case for that Kind of project, are presented. The analysis was made with input provided by the manufacturer. (Author) [pt

DRAGON 3.05D, Reactor Cell Calculation System with Burnup

International Nuclear Information System (INIS)

2007-01-01

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

TRIGA reactor main systems

International Nuclear Information System (INIS)

Boeck, H.; Villa, M.

2007-01-01

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

Status of French reactors

International Nuclear Information System (INIS)

Ballagny, A.

1997-01-01

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

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

System Design of a Supercritical CO_2 cooled Micro Modular Reactor

International Nuclear Information System (INIS)

Kim, Seong Gu; Cho, Seongkuk; Yu, Hwanyeal; Kim, Yonghee; Jeong, Yong Hoon; Lee, Jeong Ik

2014-01-01

Small modular reactor (SMR) systems that have advantages of little initial capital cost and small restriction on construction site are being developed by many research organizations around the world. Existing SMR concepts have the same objective: to achieve compact size and a long life core. Most of small modular reactors have much smaller size than the large nuclear power plant. However, existing SMR concepts are not fully modularized. This paper suggests a complete modular reactor with an innovative concept for reactor cooling by using a supercritical carbon dioxide. The authors propose the supercritical CO_2 Brayton cycle (S-CO_2 cycle) as a power conversion system to achieve small volume of power conversion unit (PCU) and to contain the reactor core and PCU in one vessel. A conceptual design of the proposed small modular reactor was developed, which is named as KAIST Micro Modular Reactor (MMR). The supercritical CO_2 Brayton cycle for the S-CO_2 cooled reactor core was optimized and the size of turbomachinery and heat exchanger were estimated preliminary. The nuclear fuel composed with UN was proposed and the core lifetime was obtained from a burnup versus reactivity calculation. Furthermore, a system layout with fully passive safety systems for both normal operation and emergency operation was proposed. (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 non-destructive methods for qualification of the U3O8-Al and U3Si2-Al dispersion fuels in the IEA-R1 Reactor

International Nuclear Information System (INIS)

Silva, Jose Eduardo Rosa da

2011-01-01

IPEN/CNEN-SP manufactures fuels to be used in its nuclear research reactor - the IEA-R1. To qualify those fuels, it is necessary to check if they have a good performance under irradiation. As Brazil doesn't have nuclear research reactors with high neutron fluxes, or suitable hot cells for carrying out post-irradiation examination of nuclear fuels, IPEN/CNEN-SP has conducted a fuel qualification program based on the use of uranium compounds, internationally tested and qualified to be used in research reactors, and has gotten experience in the technological development stages for the manufacturing of fuel plates, irradiation and non-destructive post-irradiation testing. Fuel elements containing low volume fractions of fuel in the dispersion were manufactured and irradiated successfully directly in the core of the IEA-R1. However, there are plans to increase the uranium density of these fuels. The objective of this thesis work was to study and to propose a set of non-destructive methods to qualify the dispersions fuels U 3 O 8 -Al e U 3 Si 2 -Al with high uranium density produced at IPEN/CNEN-SP. For that, the irradiation resources in the IEA-R1, and the application of non-destructive methods in the reactor pool available in the Institution were considered. The proposal is to specify, manufacture and irradiate fuel mini plates in IEA-R1 at the maximum densities, qualified internationally, and to monitor their general conditions during the period of irradiation, using non-destructive methods in the reactor pool. In addition to the non-destructive visual inspection and sipping methods, already used at the Institution, the infrastructure for dimensional sub-aquatic testing to evaluate the swelling of irradiated fuel mini plates was completed. The analyses of the results will provide means to assess and decide whether or not to continue with the irradiation of mini plates, until the desired burnup for the irradiation tests at IEA-R1 are reached. (author)

The SLOWPOKE-2 reactor with low enrichment uranium oxide fuel

International Nuclear Information System (INIS)

Townes, B.M.; Hilborn, J.W.

1985-06-01

A SLOWPOKE-2 reactor core contains less than 1 kg of highly enriched uranium (HEU) and the proliferation risk is very low. However, to overcome proliferation concerns a new low enrichment uranium (LEU) fuelled reactor core has been designed. This core contains approximately 180 fuel elements based on the Zircaloy-4 clad UOsub(2) CANDU fuel element, but with a smaller outside diameter. The physics characteristics of this new reactor core ensure the inherent safety of the reactor under all conceivable conditions and thus the basic SLOWPOKE safety philosophy which permits unattended operation is not affected

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.

VENUS-2, Reactor Kinetics with Feedback, 2-D LMFBR Disassembly Excursions

International Nuclear Information System (INIS)

Jackson, J.F.; Nicholson, R.B.; Weber, D.P.

1980-01-01

1 - Description of problem or function: VENUS-2 is an improved edition of the VENUS fast-reactor disassembly program. It is a two- dimensional (r-z) coupled neutronics-hydrodynamics code that calculates the dynamic behavior of an LMFBR during a prompt-critical disassembly excursion. It calculates the power history and fission energy release as well as the space-time histories of the fuel temperatures, core material pressures, and core material motions. Reactivity feedback effects due to Doppler broadening and reactor material motion are taken into account. 2 - Method of solution: The power and energy release are calculated using a point-kinetics formulation with up to six delayed neutron groups. The reactivity is a combination of an input driving function and feedback effects due to Doppler broadening and material motion. An adiabatic model is used to calculate the temperature increase throughout the reactor based on an initial temperature distribution and power profile provided as input data. These temperatures are, in turn, converted to fuel pressures through one of several equation of state options provided. The material motion that results from the pressure buildup is calculated by a direct finite difference solution of a set of two-dimensional (r-z) hydrodynamics equations. This is done in Lagrangian coordinates. The reactivity change associated with this motion is calculated by first-order perturbation theory. The displacements are also used to adjust the fuel densities as required for the density dependent equation-of- state option. An automatic time-step-size selection scheme is provided. 3 - Restrictions on the complexity of the problem: VENUS-2 is written so that the dimensions of the storage arrays can be readily changed to accommodate a broad range of problem sizes. In the base version, the total number of mesh intervals is restricted such that (NR+3)*(NZ+3) is less than 700, where NR and NZ are the total number of mesh intervals in the r and z

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

International Nuclear Information System (INIS)

Rosli Darmawan; Mariah Adam

2012-01-01

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

Operation of the SLOWPOKE-2 reactor in Jamaica

Energy Technology Data Exchange (ETDEWEB)

Grant, C.N.; Lalor, G.C.; Vuchkov, M.K. [University of the West Indies, Kingston (Jamaica)

2001-07-01

Over the past sixteen years lCENS has operated a SLOWPOKE 2 nuclear reactor almost exclusively for the purpose of neutron activation analysis. During this period we have adopted a strategy of minimum irradiation times while optimizing our output in an effort to increase the lifetime of the reactor core and to maintaining fuel integrity. An inter-comparison study with results obtained with a much larger reactor at IPEN has validated this approach. The parameters routinely monitored at ICENS are also discussed and the method used to predict the next shim adjustment. (author)

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

International Nuclear Information System (INIS)

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

1991-11-01

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

Irradiation techniques at BR2 reactor

International Nuclear Information System (INIS)

Hebel, W.

1978-01-01

Since 1963 the material testing reactor BR2 at Mol is operated for the realisation of numerous research programs and experiments on the behavior of materials under nuclear radiation and in particular under intensive neutron exposure. During this period special irradiation techniques and experimental devices were developed according to the desiderata of the different experiments and to the irradiation possibilities offered at BR2. The design and the operating characteristics of quite a number of those irradiation rigs of proven reliability may be used or can be made available for new irradiation experiments. A brief description is given of some typical irradiation devices designed and constructed by CEN/SCK, Technology and Energy Dpt. They are compiled according to their main use for the different research and development programs realized at BR2. Their eventual application however for different objectives could be possible. A final chapter summarizes the principal irradiation conditions offered by BR2 reactor. (author)

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)

Alteration of installation of reactors (alteration of No.1 and No.2 reactor facilities) in Oi Power Station, Kansai Electric Power Co., Inc

International Nuclear Information System (INIS)

1984-01-01

The Nuclear Safety Commission reported to the Minister of International Trade and Industry on October 27, 1983, that the technical capability was recognized to be adequate, and the safety after the alteration of the installation of reactors was judged to be ensured. At the time of deliberation, the guidelines for examining the safety design and safety evaluation of LWR facilities for power generation were used. Regarding the change of the degree of enrichment of replacement fuel from 3.2 to 3.4 wt.%, the limiting conditions are satisfied in the replacement core, and the nuclear design is appropriate. Eight test fuel assemblies using UO 2 pellets containing gadolinia are charged in the core of No.2 reactor, and the irradiation of two cycles is carried out. As the result of the safety examination regarding this test, the propriety of the nuclear design and mechanical design of the test fuel assemblies was confirmed. This alteration does not exert influence on the result of safety analysis made so far. This report was decided by the Committee on Examination of Reactor Safety based on the conclusion of No.26 subcommittee. (Kako, I.)

Chemistry in water reactors: operating experience and new developments. 2 volumes

International Nuclear Information System (INIS)

1994-01-01

These proceedings of the International conference on chemistry in water reactors (Operating experience and new developments), Volume 1, are divided into 8 sessions bearing on: (session 1) Primary coolant activity, corrosion products (5 conferences), (session 2) Dose reduction (4 conferences), (session 3) New developments (4 conferences), poster session: Primary coolant chemistry (16 posters), (session 4) Decontamination (5 conferences), poster session (2 posters), (session 5) BWR-Operating experience (3 conferences), (session 6) BWR-Modelling of operating experience (4 conferences), (session 7) BWR-Basic studies (4 conferences), (session 8) BWR-New technologies (3 conferences)

Advanced CANDU reactor pre-licensing progress

International Nuclear Information System (INIS)

Popov, N.K.; West, J.; Snell, V.G.; Ion, R.; Archinoff, G.; Xu, C.

2005-01-01

The Advanced CANDU Reactor (ACR) is an evolutionary advancement of the current CANDU 6 reactor, aimed at producing electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The Canadian Nuclear Safety Commission (CNSC) staff are currently reviewing the ACR design to determine whether, in their opinion, there are any fundamental barriers that would prevent the licensing of the design in Canada. This CNSC licensability review will not constitute a licence, but is expected to reduce regulatory risk. The CNSC pre-licensing review started in September 2003, and was focused on identifying topics and issues for ACR-700 that will require a more detailed review. CNSC staff reviewed about 120 reports, and issued to AECL 65 packages of questions and comments. Currently CNSC staff is reviewing AECL responses to all packages of comments. AECL has recently refocused the design efforts to the ACR-1000, which is a larger version of the ACR design. During the remainder of the pre-licensing review, the CNSC review will be focused on the ACR-1000. AECL Technologies Inc. (AECLT), a wholly-owned US subsidiary of AECL, is engaged in a pre-application process for the ACR-700 with the US Nuclear Regulatory Commission (USNRC) to identify and resolve major issues prior to entering a formal process to obtain standard design certification. To date, the USNRC has produced a Pre-Application Safety Assessment Report (PASAR), which contains their reviews of key focus topics. During the remainder of the pre-application phase, AECLT will address the issues identified in the PASAR. Pursuant to the bilateral agreement between AECL and the Chinese nuclear regulator, the National Nuclear Safety Administration (NNSA) and its Nuclear Safety Center (NSC), NNSA/NSC are reviewing the ACR in seven focus areas. The review started in September 2004, and will take three years. The main objective of the review is to determine how the ACR complies

Second revised edition of JENDL-3 (JENDL-3,2)

International Nuclear Information System (INIS)

Kikuchi, Yasuyuki

1995-01-01

In Nuclear Data Center, Japan Atomic Energy Research Institute, the development of Japanese standard evaluated nuclear data library, JENDL, since the beginning of 1970s has been advanced. In June, 1994, the second revised edition of the third edition, JENDL-3,2, was opened to public. This JENDL-3,2 is expected to be utilized widely in Japan and abroad by freezing it for the time being as the final version of JENDL. This report is the contents of the publication of JENDL-3,2 of the title 'Present state of Japanese nuclear data activities' made at Physical Society of Japan in 1994. Nuclear data are divided into nuclear reaction data and nuclear structure data. The nuclear reaction data on neutron induction, evaluated nuclear data file and the international cooperation for nuclear data are described. The history of Japanese nuclear data activities, nuclear data libraries in the world and the history of JENDL are introduced. As for the evaluation of JENDL-3, its outline, light, medium and heavy nuclei, the nuclides of nuclear fission products, main actinide nuclides, transuranic elements and gamma ray generation data are reported. As for the bench mark test of JENDL-3,1, its outline, simple system bench mark, fast reactor system, thermal neutron reactor system, shield bench mark and feedback information from bench mark test are described. The evaluation and the bench mark test for LENDL-3,2 are reported. (K.I.)

A regression approach for Zircaloy-2 in-reactor creep constitutive equations

International Nuclear Information System (INIS)

Yung Liu, Y.; Bement, A.L.

1977-01-01

In this paper the methodology of multiple regressions as applied to Zircaloy-2 in-reactor creep data analysis and construction of constitutive equation are illustrated. While the resulting constitutive equation can be used in creep analysis of in-reactor Zircaloy structural components, the methodology itself is entirely general and can be applied to any creep data analysis. The promising aspects of multiple regression creep data analysis are briefly outlined as follows: (1) When there are more than one variable involved, there is no need to make the assumption that each variable affects the response independently. No separate normalizations are required either and the estimation of parameters is obtained by solving many simultaneous equations. The number of simultaneous equations is equal to the number of data sets. (2) Regression statistics such as R 2 - and F-statistics provide measures of the significance of regression creep equation in correlating the overall data. The relative weights of each variable on the response can also be obtained. (3) Special regression techniques such as step-wise, ridge, and robust regressions and residual plots, etc., provide diagnostic tools for model selections. Multiple regression analysis performed on a set of carefully selected Zircaloy-2 in-reactor creep data leads to a model which provides excellent correlations for the data. (Auth.)

Status of French reactors

Energy Technology Data Exchange (ETDEWEB)

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

1997-08-01

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

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

International Nuclear Information System (INIS)

Longoni, G.

2010-01-01

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

Low-enrichment and long-life Scalable LIquid Metal cooled small Modular (SLIMM-1.2) reactor

Energy Technology Data Exchange (ETDEWEB)

El-Genk, Mohamed S., E-mail: mgenk@unm.edu [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Department, University of New Mexico, Albuquerque, NM (United States); Chemical and Biological Engineering Department, University of New Mexico, Albuquerque, NM (United States); Palomino, Luis M.; Schriener, Timothy M. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States)

2017-05-15

Highlights: • Developed low enrichment and natural circulation cooled SLIMM-1.2 SMR for generating 10–100 MW{sub th}. • Neutronics analyses estimate operation life and temperature reactivity feedback. • At 100 MW{sub th}, SLIMM-1.2 operates for 6.3 FPY without refueling. • SLIMM-1.2 has relatively low power peaking and maximum UN fuel temperature < 1400 K. - Abstract: The Scalable LIquid Metal cooled small Modular (SLIMM-1.0) reactor with uranium nitride fuel enrichment of 17.65% had been developed for generating 10–100 MW{sub th} continuously, without refueling for ∼66 and 5.9 full power years, respectively. Natural circulation of in-vessel liquid sodium (Na) cools the core of this fast energy spectrum reactor during nominal operation and after shutdown, with the aid of a tall chimney and an annular Na/Na heat exchanger (HEX) of concentric helically coiled tubes. The HEX at the top of the downcomer maximizes the static pressure head for natural circulation. In addition to the independent emergency shutdown (RSS) and reactor control (RC), the core negative temperature reactivity feedback safely decreases the reactor thermal power, following modest increases in the temperatures of UN fuel and in-vessel liquid sodium. The decay heat is removed from the core by natural circulation of in-vessel liquid sodium, with aid of the liquid metal heat pipes laid along the reactor vessel wall, and the passive backup cooling system (BCS) using natural circulation of ambient air along the outer surface of the guard vessel wall. This paper investigates modifying the SLIMM-1.0 reactor design to lower the UN fuel enrichment. To arrive at a final reactor design (SLIMM-1.2), the performed neutronics and reactivity depletion analyses examined the effects of various design and material choices on both the cold-clean and the hot-clean excess reactivity, the reactivity shutdown margin, the full power operation life at 100 MW{sub th}, the fissile production and depletion, the

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

International Nuclear Information System (INIS)

Okamoto, Koji

2015-01-01

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

Fusion reactor materials program plan. Section 2. Damage analysis and fundamental studies

International Nuclear Information System (INIS)

1978-07-01

The scope of this program includes: (1) Development of procedures for characterizing neutron environments of test facilities and fusion reactors, (2) Theoretical and experimental investigations of the influence of irradiation environment on damage production, damage microstructure evolution, and mechanical and physical property changes, (3) Identification and, where appropriate, development of essential nuclear and materials data, and (4) Development of a methodology, based on damage mechanisms, for correlating the mechanical behavior of materials exposed to diverse test environments and projecting this behavior to magnetic fusion reactor (MFR) environments. Some major problem areas are addressed

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

International Nuclear Information System (INIS)

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

2008-01-01

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

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

TMI-2 reactor vessel head removal

International Nuclear Information System (INIS)

Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

1985-09-01

This report describes the safe removal and storage of the Three Mile Island Unit 2 (TMI-2) reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training, the head was safely removed and stored; and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

TMI-2 reactor vessel plenum final lift

International Nuclear Information System (INIS)

Wilson, D.C.

1986-01-01

Removal of the plenum assembly from the TMI-2 reactor vessel was necessary to gain access to the core region for defueling. The plenum was lifted from the reactor vessel by the polar crane using three specially designed pendant assemblies. It was then transferred in air to the flooded deep end of the refueling canal and lowered onto a storage stand where it will remain throughout the defueling effort. The lift and transfer were successfully accomplished on May 15, 1985 in just under three hours by a lift team located in a shielded area within the reactor building. The success of the program is attributed to extensive mockup and training activities plus thorough preparations to address potential problems. 54 refs

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

TOKMINA, Toroidal Magnetic Field Minimization for Tokamak Fusion Reactor. TOKMINA-2, Total Power for Tokamak Fusion Reactor

International Nuclear Information System (INIS)

Hatch, A.J.

1975-01-01

1 - Description of problem or function: TOKMINA finds the minimum magnetic field, Bm, required at the toroidal coil of a Tokamak type fusion reactor when the input is beta(ratio of plasma pressure to magnetic pressure), q(Kruskal-Shafranov plasma stability factor), and y(ratio of plasma radius to vacuum wall radius: rp/rw) and arrays of PT (total thermal power from both d-t and tritium breeding reactions), Pw (wall loading or power flux) and TB (thickness of blanket), following the method of Golovin, et al. TOKMINA2 finds the total power, PT, of such a fusion reactor, given a specified magnetic field, Bm, at the toroidal coil. 2 - Method of solution: TOKMINA: the aspect ratio(a) is minimized, giving a minimum value for Bm. TOKMINA2: a search is made for PT; the value of PT which minimizes Bm to the required value within 50 Gauss is chosen. 3 - Restrictions on the complexity of the problem: Input arrays presently are dimensioned at 20. This restriction can be overcome by changing a dimension card

On-line reactor building integrity testing at Gentilly-2 (summary of results 1987-1994)

International Nuclear Information System (INIS)

Collins, N.; Lafreniere, P.

1994-01-01

In 1987, Hydro-0uebec embarked on an ambitious development program to provide the Gentilly-2 Nuclear Power Station with an effective and practical Reactor Building Containment integrity Test (CIT). In October 1992, the inaugural low pressure (3 kPa(g) nominal) CIT at 100% F.P was performed. The test was conclusive and the CIT was declared In-Service for containment integrity verification on-line. Five subsequent CITs performed in 1993 and 1994 have demonstrated the expected leak rate results and good reliability. The outstanding feature of the CITs is the demonstrated accurary of better than 5% of the measured leak rate. The CIT was developed with the primary goal of demonstrating 'overall' containment availability. Specifically it was designed to detect a 25 mm. diameter leak or hole in the Reactor Building. However, the remarkable CIT accuracy allows reliable detection of a 2 mm. hole. The Gentilly-2 CIT is an innovative approach based on the Temperature Compensation Method (TCM) which uses a reference volume composed of an extensive tubular network of several different diameters. This eliminates the need to track numerous temperature points. A second independent tubular network includes numerous humidity sampling points, thereby enabling the mearurernent of minute pressure variations inside the Reactor Building, independant of the spatial and temporal humidity behaviour. This Gentilly-2 TOM System has been demonstrated to work at both high and low test pressures. The GentiIly-2 design allows the CIT to be performed at a nominal 3 kPa(g) test pressure during a 12-hour period (28 hours total with alignment time) with the reactor at full power. The traditional Reactor Building Pressure Test (RBPT) is typically performed at high pressure (124 kPa(g) in a 5-day critical path window (7 days total with alignment time) during an annual shutdown

MASTER-2.0: Multi-purpose analyzer for static and transient effects of reactors

Energy Technology Data Exchange (ETDEWEB)

Cho, Byung Oh; Song, Jae Seung; Joo, Han Gyu [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-01-01

MASTER-2.0 (Multi-purpose Analyzer for Static and Transient Effects of Reactors) is a nuclear design code based on the two group diffusion theory to calculate the steady-state and transient pressurized water reactor core in a 3-dimensional Cartesian or hexagonal geometry. Its neutronics model solves the space-time dependent neutron diffusion equations with NIM(Nodal Integration Method), NEM (Nodal Expansion Method), AFEN (Analytic Function Expansion Nodal Method)/NEM Hybrid Method, NNEM (Non-linear Nodal Expansion Method) or NANM (Non-linear Analytic Nodal Method) for a Cartesian geometry and with AFEN/NEM Hybrid Method or NLFM (Non-linear Local Fine-Mesh Method) for a hexagonal one. Coarse mesh rebalancing, Krylov Subspace method and asymptotic extrapolation method are implemented to accelerate the convergence of iteration process. Master-2.0 performs microscopic depletion calculations using microscopic cross sections provided by CASMO-3 or HELIOS and also has the reconstruction capability of pin information by use of MSS-IAS (Method of Successive Smoothing with Improved Analytic Solution). For the thermal-hydraulic calculation, fuel temperature table or COBRA3-C/P model can be used selectively. In addition, MASTER-2.0 is designed to cover various PWRs including SMART as well as WH-and CE-type reactors, providing all data required in their design procedures. (author). 39 refs., 12 figs., 4 tabs.

Refurbishment programme for the BR2-reactor

Energy Technology Data Exchange (ETDEWEB)

Koonen, E [Centre d' Etude de l' Energie Nucleaire, Studiecentrum voor Kernenergie, BR2 Department, Boeretang, Mol (Belgium)

1992-07-01

BR2 is a high flux engineering test reactor, which differs from comparable material testing reactors by its specific core array (fig. 1). It is a heterogeneous, thermal, tank-in-pool type reactor, moderated by beryllium and light water, which serves also as coolant. The fuel elements consist of cylindrical assemblies loaded in channels materialized by hexagonal beryllium prisms. The central 200 mm channel is vertical, while all others are inclined and form a hyperbolical arrangement around the central one. This feature combines a very compact core with the requirement of sufficient space for individual access to all channels through penetrations in the top cover of the aluminium pressure vessel. Each channel may hold a fuel element, a control rod, an experiment, an irradiation device or a beryllium plug. The refurbishment Program According to the present programme of C.E.N./S.C.K., BR2 will be in operation until 1996. At that time, the beryllium matrix will reach its foreseen end-of-life. In order to continue operation beyond this point, a thorough refurbishment of the reactor is foreseen, in addition to the unavoidable replacement of the matrix, to ensure quality of the installation and compliance with modern standards. Some fundamental options have been taken as a starting point: BR2 will continue to be used as a classical MTR, i.e. fuel and material irradiations and safety experiments with some additional service-activities. The present configuration is optimized for that use and there is no specific experimental requirement to change the basic concepts and performance characteristics. From the customers viewpoint, it is desirable to go ahead with the well-known features of BR2, to maintain a high degree of availability and reliability and to minimize the duration of the long shutdown. It is also important to limit the amount of nuclear liabilities. So the objective of the refurbishment programme is the life extension of BR2 for about 15 years, corresponding to

Refurbishment programme for the BR2-reactor

International Nuclear Information System (INIS)

Koonen, E.

1992-01-01

BR2 is a high flux engineering test reactor, which differs from comparable material testing reactors by its specific core array (fig. 1). It is a heterogeneous, thermal, tank-in-pool type reactor, moderated by beryllium and light water, which serves also as coolant. The fuel elements consist of cylindrical assemblies loaded in channels materialized by hexagonal beryllium prisms. The central 200 mm channel is vertical, while all others are inclined and form a hyperbolical arrangement around the central one. This feature combines a very compact core with the requirement of sufficient space for individual access to all channels through penetrations in the top cover of the aluminium pressure vessel. Each channel may hold a fuel element, a control rod, an experiment, an irradiation device or a beryllium plug. The refurbishment Program According to the present programme of C.E.N./S.C.K., BR2 will be in operation until 1996. At that time, the beryllium matrix will reach its foreseen end-of-life. In order to continue operation beyond this point, a thorough refurbishment of the reactor is foreseen, in addition to the unavoidable replacement of the matrix, to ensure quality of the installation and compliance with modern standards. Some fundamental options have been taken as a starting point: BR2 will continue to be used as a classical MTR, i.e. fuel and material irradiations and safety experiments with some additional service-activities. The present configuration is optimized for that use and there is no specific experimental requirement to change the basic concepts and performance characteristics. From the customers viewpoint, it is desirable to go ahead with the well-known features of BR2, to maintain a high degree of availability and reliability and to minimize the duration of the long shutdown. It is also important to limit the amount of nuclear liabilities. So the objective of the refurbishment programme is the life extension of BR2 for about 15 years, corresponding to

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

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

International Nuclear Information System (INIS)

Kuscu, Ozlem Selcuk; Sponza, Delia Teresa

2009-01-01

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

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

Impact of 2D/3D-project on LOCA-licensing analysis and reactor safety of PWRs

International Nuclear Information System (INIS)

Winkler, F.; Krebs, W.D.

1989-01-01

In the past LOCA-licensing analysis has included large conservatisms to compensate for the lack of detailed two phase flow and full scale experimental data. The 2D/3D-project was established to improve the data base in order to minimize the conservatisms required. The significant results and findings of the full scale Upper Plenum Test Facility (UPTF) and from the electrically heated Slab Core Test Facility (SCTF) were particularly useful for understanding the multidimensional phenomena in the primary system and in the core of a PWR. UPTF results were used to verify the TRAC-PF1 analysis of a PWR with combined ECC-Injection during the reflood phase of a large break-LOCA. Comparison of these results with results from classic licensing calculations quantifies the large safety margin in earlier licensing procedures and in reactor systems. (orig.)

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

Energy Technology Data Exchange (ETDEWEB)

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

1993-11-01

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

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

International Nuclear Information System (INIS)

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

1982-03-01

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

Systems analysis of the CANDU 3 Reactor

Energy Technology Data Exchange (ETDEWEB)

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

1993-07-01

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

Nuclear safety risk control in the outage of CANDU unit

International Nuclear Information System (INIS)

Wu Mingliang; Zheng Jianhua

2014-01-01