Experience of oil in CANDU moderator during A831 planned outage at Bruce Power

International Nuclear Information System (INIS)

Ma, G.; Nashiem, R.; Matheson, S.; Stuart, C.; Roberts, J.G.

2011-01-01

In their address to the Nuclear Plant Chemistry Conference 2009, Bruce Power staff will describe the effects of oil ingress to the moderator of a CANDU reactor. During the A831 planned outage of Bruce Power Unit 3, an incident of oil ingress into moderator was discovered on Oct 17, 2008. An investigation identified the cause of the oil ingress. Atomic Energy of Canada Ltd. (AECL) assessed operability of the reactor with the oil present and made recommendations with respect to the effect on unit start-up with oil present. The principal concern was the radiolytic generation of deuterium from the breakdown of the oil in-core. Various challenges were presented during start-up which were overcome via innovative approaches. The subsequent actions and consequential effects on moderator chemistry are discussed in this paper. Examination of the plant chemistry data revealed some interesting aspects of moderator system chemistry under upset conditions which will also be presented. (author)

Experience of oil in CANDU® moderator during A831 planned outage at Bruce Power

International Nuclear Information System (INIS)

Ma, G.; Nashiem, R.; Matheson, S.; Stuart, C.; Roberts, J.G.

2010-01-01

In their address to the Nuclear Plant Chemistry Conference 2009, Bruce Power staff will describe the effects of oil ingress to the moderator of a CANDU® reactor. During the A831 planned outage of Bruce Power Unit 3, an incident of oil ingress into moderator was discovered on Oct 17, 2008. An investigation identified the cause of the oil ingress. Atomic Energy of Canada Ltd. (AECL) assessed operability of the reactor with the oil present and made recommendations with respect to the effect on unit start-up with oil present. The principal concern was the radiolytic generation of deuterium from the breakdown of the oil in-core. Various challenges were presented during start-up which were overcome via innovative approaches. The subsequent actions and consequential effects on moderator chemistry are discussed in this paper. Examination of the plant chemistry data revealed some interesting aspects of moderator system chemistry under upset conditions which will also be presented. (author)

Experience of oil in CANDU® moderator during A831 planned outage at Bruce Power

Energy Technology Data Exchange (ETDEWEB)

Ma, G.; Nashiem, R.; Matheson, S. [Bruce Power, Tiverton, Ontario (Canada); Stuart, C. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Roberts, J.G. [CANTECH Associates Ltd., Burlington, Ontario (Canada)

2010-07-01

In their address to the Nuclear Plant Chemistry Conference 2009, Bruce Power staff will describe the effects of oil ingress to the moderator of a CANDU® reactor. During the A831 planned outage of Bruce Power Unit 3, an incident of oil ingress into moderator was discovered on Oct 17, 2008. An investigation identified the cause of the oil ingress. Atomic Energy of Canada Ltd. (AECL) assessed operability of the reactor with the oil present and made recommendations with respect to the effect on unit start-up with oil present. The principal concern was the radiolytic generation of deuterium from the breakdown of the oil in-core. Various challenges were presented during start-up which were overcome via innovative approaches. The subsequent actions and consequential effects on moderator chemistry are discussed in this paper. Examination of the plant chemistry data revealed some interesting aspects of moderator system chemistry under upset conditions which will also be presented. (author)

Experience of oil in CANDU moderator during A831 planned outage at Bruce Power

Energy Technology Data Exchange (ETDEWEB)

Ma, G.; Nashiem, R.; Matheson, S. [Bruce Power, Tiverton, Ontario (Canada); Stuart, C. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Roberts, J.G. [CANTECH Associates Ltd., Burlington, Ontario (Canada)

2011-03-15

In their address to the Nuclear Plant Chemistry Conference 2009, Bruce Power staff will describe the effects of oil ingress to the moderator of a CANDU reactor. During the A831 planned outage of Bruce Power Unit 3, an incident of oil ingress into moderator was discovered on Oct 17, 2008. An investigation identified the cause of the oil ingress. Atomic Energy of Canada Ltd. (AECL) assessed operability of the reactor with the oil present and made recommendations with respect to the effect on unit start-up with oil present. The principal concern was the radiolytic generation of deuterium from the breakdown of the oil in-core. Various challenges were presented during start-up which were overcome via innovative approaches. The subsequent actions and consequential effects on moderator chemistry are discussed in this paper. Examination of the plant chemistry data revealed some interesting aspects of moderator system chemistry under upset conditions which will also be presented. (author)

Ion chamber repairs in Bruce A

International Nuclear Information System (INIS)

Millard, J.; Edwards, T.; Kerker, J.; Pletch, R.; Edwards, T.

2012-01-01

This paper discusses identification and successful remediation of leakage of shield tank water on vertical and horizontal Ion Chambers in Bruce A. In doing so, it discusses real events moving from the initial investigation to understand the problem, through looking at options for solutions, and moving to site work and actual resolution.. In multiunit 900 MW class CANDU® reactors, the calandria vessel is suspended within a larger shield tank. Due to temperature changes or changes in moderator fluid levels in the calandria, the calandria can move relative to the shield tank and its reactivity deck. Thimbles which contain the reactivity sensors and controls connect the two vessels and allow the reactivity drives and controls connections to be placed on the deck structure on the top of the reactor assembly for RRS and SDS1 and horizontally for SDS2. These thimbles have expansion joints with metal bellows where they meet the deck structure or shield tank walls. The deck structure lies on a vault containment boundary. The horizontal ion chambers are not in the containment boundary as they connect the outside of the calandria and shield tank around mid plane in the reactor vault, but due to geometry difference provides a more challenging work environment. Bruce had a beetle alarm (1-63851-MIA2-ME30 in alarm state (vertical IC housing)) at the start of April 2012 on Unit 1 channel F vertical Ion chamber expansion joint at the deck connection. This occurred after the moderator levels had been raised after the several years long refurbishment outage and the expansion joint had a significant travel. The investigation showed shield tank water in the collection chamber at the beetle. In addition, Channel J of the horizontal ion chamber had a seized instrument, which on removal was found to relate to oxide build up as a result of minor water leakage into the site. Repairs in both cases were performed as part of the long Bruce 1 & 2 refurbishment outage to completely stop the

Bruce Power - the first 24 exciting months

International Nuclear Information System (INIS)

Mottram, R.

2003-01-01

In this presentation, Ron Mottram will review the 2 year business evolution since the inception of Bruce Power - Ontario's largest independent electricity generator. Mr. Mottram will provide an overview of the Bruce Power business and operating history, along with specific emphasis on the project to Restart Bruce Units 3 and 4. Ron will share many of the project successes and challenges, and will provide insight into the myriad of issues faced by a large multi-faceted project of this type. (author)

RCM: the Bruce B experience

International Nuclear Information System (INIS)

Hill, Earl S.; Doyle, E.K.

1995-01-01

The use of RCM techniques have begun to change maintenance practice at Bruce B. This paper identifies the status of the program at Bruce B, and examines a new methodology for completing system analysis studies by incorporating lessons learned and results from Bruce A. (author)

AECB staff review of Bruce NGS 'B' operation for the year 1989

International Nuclear Information System (INIS)

1990-06-01

The operation of the Bruce Nuclear Generating Station 'B' is monitored and licensing requirements are enforced by the Atomic Energy Control Board (AECB), which observes operation of the reactors, conducts audits, witnesses important activities, reviews station documentation and reports, and issue approvals, where appropriate, in accordance with licence conditions. This report records the conclusions of the AECB staff assessment of Bruce NGS 'B' during 1989. In general, the station operated within acceptable safety standards. Quality improvement initiatives started in 1989 should lead to improved station maintenance and operation in coming years. Ontario Hydro still needs to improve the administration of operating memos, deficiency reports and call-ups. Station management must ensure that shift supervisors and reactor first operators operate the station in a conservative manner at all times and put safety interests first when responding to a unit upset. (2 tabs.)

AECB staff annual report of Bruce NGS 'B' for the year 1988

International Nuclear Information System (INIS)

1989-05-01

The operation of the Bruce 'B' Nuclear Generating Station is monitored and licensing requirements are enforced by the Atomic Energy Control Board (AECB) Bruce project staff, with appropriate support from other AECB personnel. The staff observes operation of the reactors, conducts audits, witnesses important activities, reviews station documentation and reports, and issues approvals where appropriate in accordance with license conditions. As required by a condition of its Operating Licence, Ontario Hydro each year submits Technical Reports which summarize various aspects of the operation of Bruce NGS 'B' during the year. When these reports have been reviewed by AECB staff, a formal Annual Review Meeting is held with the station management to discuss safety-related aspects of the station operation, and to inform Ontario Hydro of AECB staff conclusions with respect to the performance of Ontario Hydro in operating the station during the year

Bruce A units 1 and 2 restart project

International Nuclear Information System (INIS)

Routledge, K.

2006-01-01

This presentation provides an overview of the Bruce A Units 1 and 2 Restart project from the vantage point of the Project Management Contractor (PMC). The presentation will highlight the unique structure of the project, which has been designed to maximize project efficiencies while minimizing the impact to the Bruce Power operational reactors. Efficiency improvements covered in the presentation includes: support services provided to the direct work contractors, radiation protection, worker protection, engineering, field execution, maintenance and facilities. The presentation focusses on the roles of the PMC in helping to ensure the successful outcome of this ambitious reactor refurbishment project. In addition, the Construction Island concept that has been implemented on the project will be presented, with some of the innovative thinking that has gone into its creation. The organization of the PMC and an overview of the project schedule is also presented. AMEC NCL is a privately held consultancy in the Canadian nuclear industry which provides experienced and flexible multi-disciplined resources to support full project management, engineering solutions and safety consultancy services throughout the life cycle of nuclear facilities in Canada, and for customers in related markets in North America and overseas. AMEC NCL is a wholly-owned subsidiary of AMEC plc

AECB staff annual assessment of the Bruce A Nuclear Generating Station for the year 1996

International Nuclear Information System (INIS)

1997-06-01

The Atomic Energy Control Board is the independent federal agency that controls all nuclear activities in Canada. A major use of nuclear energy in Canada is electricity production. The AECB assesses every station's performance against legal requirements, including the conditions in the operating licence. Each station is inspected and all aspects of the station's operation and management is reviewed. This report is the AECB staff assessment of reactor safety at the Bruce Nuclear Generating Station A for 1996. Ontario Hydro operated Bruce A safely in 1996, maintaining the risk to workers and the public at an acceptably low level. Special safety system performance at Bruce A was adequate. Availability targets were all met. Improvement is needed to reduce the number of operating licence non-compliances

Bruce NGS a loss of flow analysis for effectiveness of level 2 defence-in-depth provisions

International Nuclear Information System (INIS)

Won, W.; Jiang, Y.; Kwee, M.; Xue, J.

2014-01-01

The concept of defence-in-depth is applied to CANDU (CANada Deuterium Uranium) reactor designs and operations to provide series of levels of defence to prevent accidents progressing and to provide protection for reactor and public safety. The level 2 defence-in-depth provisions are designed to detect and intercept deviation from normal operation in order to prevent anticipated operating occurrences (AOOs) from escalating to accident conditions, and to return the plant to a state of normal operations, according to the Canada Nuclear Safety Commission (CNSC) regulatory document RD-337. Historically, safety analysis has focused on the effectiveness of level 3 defence-in-depth provisions in accident conditions, and the effectiveness of level 2 defence-in-depth has not been assessed. In this study, the effectiveness of Level 2 defence-in-depth is assessed for loss of flow (LOF) events for Bruce Nuclear Generating Station (NGS) A reactors. The level 2 defence-in-depth in Bruce NGS A design is identified to be the stepback function of reactor regulating system (RRS). The behavior of RRS stepback following the initiation of loss of flow event is simulated using RFSP/TUF/RRS - em coupled code. The behavior of full system and single channel is simulated and assessed against the acceptance criteria - fitness for service of systems, structures and components (SSCs). (author)

Bruce NGS a loss of flow analysis for effectiveness of level 2 defence-in-depth provisions

Energy Technology Data Exchange (ETDEWEB)

Won, W. [AMEC NSS, Toronto, ON (Canada); Jiang, Y.; Kwee, M.; Xue, J. [Bruce Power, Toronto, ON (Canada)

2014-07-01

The concept of defence-in-depth is applied to CANDU (CANada Deuterium Uranium) reactor designs and operations to provide series of levels of defence to prevent accidents progressing and to provide protection for reactor and public safety. The level 2 defence-in-depth provisions are designed to detect and intercept deviation from normal operation in order to prevent anticipated operating occurrences (AOOs) from escalating to accident conditions, and to return the plant to a state of normal operations, according to the Canada Nuclear Safety Commission (CNSC) regulatory document RD-337. Historically, safety analysis has focused on the effectiveness of level 3 defence-in-depth provisions in accident conditions, and the effectiveness of level 2 defence-in-depth has not been assessed. In this study, the effectiveness of Level 2 defence-in-depth is assessed for loss of flow (LOF) events for Bruce Nuclear Generating Station (NGS) A reactors. The level 2 defence-in-depth in Bruce NGS A design is identified to be the stepback function of reactor regulating system (RRS). The behavior of RRS stepback following the initiation of loss of flow event is simulated using RFSP/TUF/RRS{sub -}em coupled code. The behavior of full system and single channel is simulated and assessed against the acceptance criteria - fitness for service of systems, structures and components (SSCs). (author)

The operation and maintenance of the SLAR system at Bruce A

Energy Technology Data Exchange (ETDEWEB)

Ahuja, S [Ontario Hydro, Tiverton, ON (Canada). Bruce Nuclear Generating Station-A

1997-12-31

The SLAR (Spacer Location And Repositioning) system at Bruce A consists of two (2) Delivery Machines and a Fuelling Machine Trolley equipped with the D{sub 2}0 and Air auxiliary systems. The Delivery Machines are designed to perform all the Fuelling Machine operations and have the capability to rapidly defuel/refuel a reactor channel and traverse the SLAR tool to locate and reposition the spacers in the channel. The number of functions that a Delivery Machine must perform makes it more complex as compared to the operations of a Fuelling Machine. The paper discusses the operation of the SLAR Delivery Machines and the problems encountered with the operation and maintenance of this system at Bruce A. (author). 8 figs.

Carsten Niebuhr and James Bruce

DEFF Research Database (Denmark)

Friis, Ib

2013-01-01

In 1791 Carsten Niebuhr published a review of the first two volumes of Bruce’s Reisen zur Entdeckung der Quellen des Nils (1790). Niebuhr’s strongest criticism of Bruce was that he seemed to have plagiarized some of Niebuhr’s astronomical observations (“adopted them without examination”) and that......In 1791 Carsten Niebuhr published a review of the first two volumes of Bruce’s Reisen zur Entdeckung der Quellen des Nils (1790). Niebuhr’s strongest criticism of Bruce was that he seemed to have plagiarized some of Niebuhr’s astronomical observations (“adopted them without examination...... as written by Bruce in 1770 at Gondar, Abyssinia, contains information about latitudes identical with some of Niebuhr’s observations which were unpublished in 1770; possible explanations for this are proposed. In summary, it seems that Niebuhr is right; it is almost certain that Bruce plagiarized some...

The Bruce nuclear project

International Nuclear Information System (INIS)

Rose, J.B.

1981-01-01

This case study assesses the industrial relations impact of the construction of the Bruce Nuclear Power Development. It examines the labour relations system in the Ontario electric power sector and in major building construction. Industrial relations problems and practices at the Bruce project are reviewed. The focus of the study is on the relationship between the project and the rest of the Ontario industrial construction industry

Ontario Hydro's operating experience with steam generators with specifics on Bruce A and Bruce B problems

Energy Technology Data Exchange (ETDEWEB)

Eatock, J W; Patterson, R W [Ontario Hydro, Toronto, ON (Canada); Dyck, R W [Ontario Hydro, Central Production Services Division, Toronto, ON (Canada)

1991-04-01

The performance of the steam generators in Ontario Hydro nuclear power stations is reviewed. This performance has generally been outstanding compared to world averages, with very low tube failure and plugging rates. Steam generator problems have made only minor contributions to Ontario Hydro nuclear station incapability factors. The mechanisms responsible for the the observed tube degradation and failures are described. The majority of the leaks have been due fatigue in the U-bend of the Bruce 'A' steam generators. There have been very few failures attributed to corrosion of the three tube materials used in Ontario Hydro steam generators. Recent performance has been deteriorating primarily due to deposit accumulation in the steam generators. Plugging of the broached holes in the upper support plates at Bruce 'A' has caused some derating of two units. Increases have been observed in the primary heat transport system reactor inlet temperature of several units. These increases may be attributed to steam generator tube surface fouling. In addition, several units have accumulated deep, hard sludge piles on the tube sheet, although little damage been observed. Recently some fretting of tubes has been observed at BNGSB in the U-bend support region. Remedial measures are being taken to address the current problems. Solutions are being evaluated to reduce the generation of corrosion products in the feedtrain and their subsequent transport to the steam generators. (author)

Ontario Hydro's operating experience with steam generators with specifics on Bruce A and Bruce B problems

International Nuclear Information System (INIS)

Eatock, J.W.; Patterson, R.W.; Dyck, R.W.

1991-01-01

The performance of the steam generators in Ontario Hydro nuclear power stations is reviewed. This performance has generally been outstanding compared to world averages, with very low tube failure and plugging rates. Steam generator problems have made only minor contributions to Ontario Hydro nuclear station incapability factors. The mechanisms responsible for the the observed tube degradation and failures are described. The majority of the leaks have been due fatigue in the U-bend of the Bruce 'A' steam generators. There have been very few failures attributed to corrosion of the three tube materials used in Ontario Hydro steam generators. Recent performance has been deteriorating primarily due to deposit accumulation in the steam generators. Plugging of the broached holes in the upper support plates at Bruce 'A' has caused some derating of two units. Increases have been observed in the primary heat transport system reactor inlet temperature of several units. These increases may be attributed to steam generator tube surface fouling. In addition, several units have accumulated deep, hard sludge piles on the tube sheet, although little damage been observed. Recently some fretting of tubes has been observed at BNGSB in the U-bend support region. Remedial measures are being taken to address the current problems. Solutions are being evaluated to reduce the generation of corrosion products in the feedtrain and their subsequent transport to the steam generators. (author)

Development and implementation of the heavy water program at Bruce Power

International Nuclear Information System (INIS)

Davloor, R.; Bourassa, C.

2014-01-01

Bruce Power operates 8 pressurized heavy water reactor units requiring more than 6000 mega grams (Mg) of heavy water. A Heavy Water Management Program that has been developed to administer this asset over the past 3 years. Through a corporate management system the Program provides governance, oversight and support to the stations. It is implemented through organizational structure, program and procedure documents and an information management system that provides benchmarked metrics, business intelligence and analytics for decision making and prediction. The program drives initiatives such as major maintenance activities, capital programs, detritiation strategies and ensures heavy water systems readiness for outages and rehabilitation of units. (author)

Development and implementation of the heavy water program at Bruce Power

Energy Technology Data Exchange (ETDEWEB)

Davloor, R.; Bourassa, C., E-mail: ram.davloor@brucepower.com, E-mail: carl.bourassa@brucepower.com [Bruce Power, Tiverton, ON (Canada)

2014-07-01

Bruce Power operates 8 pressurized heavy water reactor units requiring more than 6000 mega grams (Mg) of heavy water. A Heavy Water Management Program that has been developed to administer this asset over the past 3 years. Through a corporate management system the Program provides governance, oversight and support to the stations. It is implemented through organizational structure, program and procedure documents and an information management system that provides benchmarked metrics, business intelligence and analytics for decision making and prediction. The program drives initiatives such as major maintenance activities, capital programs, detritiation strategies and ensures heavy water systems readiness for outages and rehabilitation of units. (author)

Bruce A restart (execution and lessons-learned)

International Nuclear Information System (INIS)

Soini, J.

2011-01-01

Lessons learned with the Bruce Units 3 and 4 restart have been incorporated into the current refurbishment of Units 1 and 2. In addition, lessons learned on the lead unit (U2) are aggressively applied on the lagging unit (U1) to maximize efficiency and productivity. There will be a discussion on how this internal OPEX, along with external lessons learned, are used to continuously improve all aspects of the Bruce A Restart project management cycle, from scope selection, through planning and scheduling, to execution.

Bruce nuclear power development (BNPD) postoperational aquatic studies

International Nuclear Information System (INIS)

Carey, W.E.

1984-01-01

This report summarizes the results of three years of postoperational aquatic study conducted between 1979 and 1981 at the Bruce Nuclear Power Development site. An increase in the rate of organic and inorganic sedimentation during the summer was noted, and was possibly related to construction activity at the Bruce GS 'B' intake site. Vertical thermal stratification persisted later in the year at the 7 m contour of Bruce GS 'A' discharge than at other locations sampled. Water quality conditions reflected the oliogtrophic state of Lake Huron. Several changes were noted in the biotic community. The taxonomic composition of attached algae, zooplankton and benthic macroinvertebrates varied between sampling years. The number of common naids, amphipods and the trichopteran Cheumatopsyche increased substantially in the 1981 rock cage collections. The relative abundance of adult walleye, channel catfish and round whitefish in gill nets increased, with the former two species being more abundant (15 fish per net in September, 1980, and 33 fish per net in July, 1981, respectively) at the 3 m contour of the Bruce GS 'A' discharge transect than at other shoreline sampling locations

Business health reporting process at Bruce Power helps drive successful plant performance

International Nuclear Information System (INIS)

Krane, J.C.

2007-01-01

Developing and implementing consistent and comprehensive measures of performance on a large multi-reactor unit nuclear power plant site is a significant challenge. Linking these performance measures back to licence compliance standards and all aspects of the operations, engineering, maintenance and support activities is needed to ensure cohesive site-wide safe operations and satisfy regulatory needs. At Bruce Power, Canada's largest independently-owned nuclear power producer, a Business Health reporting process has been developed to provide a standardized performance rating scheme. The reporting process ties all self assessment activities to common management principles and process structure areas that comprise the Bruce Power Management System. The principles used for performance ratings link directly back to the operating licenses and the primary referenced management system standard. The Business Health reporting process provides a natural business and regulatory oversight framework report that is easily understood and consistently measured over time. The rating data is derived from easily understood quantitative and qualitative descriptions that can be trended over time. The results derived from semi-annual Business Health reports provide an ongoing overall measure of Bruce Power's management system effectiveness for enabling and sustaining required business results and high standards of safety. (author)

Bruce unit 1 moderator to end shield cooling leak repairs

Energy Technology Data Exchange (ETDEWEB)

Boucher, P; Ashton, A [Ontario Hydro, Tiverton, ON (Canada). Bruce Nuclear Generating Station-A

1996-12-31

In October 1994, a leak developed between the heavy water Moderator System and the light water End Shield Cooling System at Ontario Hydro`s Bruce A Generating Station Unit 1. The interface between these two systems consists of numerous reactor components all within the reactor vessel. This paper describes the initial discovery and determination of the leak source. The techniques used to pinpoint the leak location are described. The repair strategies and details are outlined. Flushing and refilling of the Moderator system are discussed. The current status of the Unit 1 End Shield Cooling System is given with possible remedial measures for clean-up. Recommendations and observations are provided for future references. (author). 7 figs.

AECB staff annual assessment of the Bruce B Nuclear Generating Station for the year 1994

International Nuclear Information System (INIS)

1995-06-01

AECB staff believes Ontario Hydro operated Bruce B safely in 1994. The Bruce B reactors will remain limited to 88% full power until Ontario Hydro is able to demonstrate that it is safe to operate at higher powers. Ontario Hydro's compliance with AECB regulations and the Operating Licence was satisfactory. AECB found no major violations. The station performance was similar to previous years. Radiation doses to workers and the public were well below the legal limits and also remained below Ontario Hydro's internal targets. Worker radiation doses increased slightly but were comparable to previous years. Inspection of pressure tubes and steam generator tubes by Ontario Hydro showed continuing tube degradation. However, we believe that Ontario Hydro made progress in correcting and managing these problems. Ontario Hydro carried out a full-scale fire drill at Bruce B in 1994. AECB witnessed the drill and were pleased to observe a significant improvement in the station's fire-fighting capability. 7 tabs., 4 figs

Impact of bundle deformation on CHF: ASSERT-PV assessment of extended burnup Bruce B bundle G85159W

International Nuclear Information System (INIS)

Rao, Y.F.; Manzer, A.M.

2005-01-01

This paper presents a subchannel thermalhydraulic analysis of the effect on critical heat flux (CHF) of bundle deformation such as element bow and diametral creep. The bundle geometry is based on the post-irradiation examination (PIE) data of a single bundle from the Bruce B Nuclear Generating Station, Bruce B bundle G85159W, which was irradiated for more than two years in the core during reactor commissioning. The subchannel code ASSERT-PV IST is used to assess changes in CHF and dryout power due to bundle deformation, compared to the reference, undeformed bundle. (author)

Fuel defect detection, localization and removal in Bruce Power units 3 through 8

International Nuclear Information System (INIS)

Stone, R.; Armstrong, J.; Iglesias, F.; Oduntan, R.; Lewis, B.

2005-01-01

Fuel element defects are occurring in Bruce 'A' and Bruce 'B' Units. A root-cause investigation is ongoing, however, a solution is not yet in-hand. Fuel defect management efforts have been undertaken, therefore, in the interim. Fuel defect management tools are in-place for all Bruce Units. These tools can be categorized as analysis-based or operations-based. Analysis-based tools include computer codes used primarily for fuel defect characterization, while operations-based tools include Unit-specific delayed-neutron ('DN') monitoring systems and gaseous fission product ('GFP') monitoring systems. Operations-based tools are used for fuel defect detection, localization and removal activities. Fuel and Physics staff use defect detection, localization and removal methodologies and guidelines to disposition fuel defects. Methodologies are 'standardized' or 'routine' procedures for implementing analysis-based and operations-based tools to disposition fuel defects during Unit start-up operation and during operation at high steady-state power levels. Guidelines at present serve to supplement fuel defect management methodologies during Unit power raise. (author)

Performance of Bruce natural UO2 fuel irradiated to extended burnups

International Nuclear Information System (INIS)

Zhou, Y.N.; Floyd, M.R.; Ryz, M.A.

1995-11-01

Bruce-type bundles XY, AAH and GF were successfully irradiated in the NRU reactor at Chalk River Laboratories to outer-element burnups of 570-900 MWh/kgU. These bundles were of the Bruce Nuclear Generating Station (NGS)-A 'first-charge' design that contained gas plenums in the outer elements. The maximum outer-element linear powers were 33-37 kW/m. Post-irradiation examination of these bundles confirmed that all the elements were intact. Bundles XY and AAH, irradiated to outer-element burnups of 570-700 MWh/kgU, experienced low fission-gas release (FGR) ( 500 MWh/kgU (equivalent to bundle-average 450 MWh/kgU) when maximum outer-element linear powers are > 50 kW/m. The analysis in this paper suggests that CANDU 37-element fuel can be successfully irradiated (low-FGR/defect-free) to burnups of at least 700 MWh/kgU, provided maximum power do not exceed 40 kW/m. (author). 5 refs., 1 tab., 8 figs

Review of the reliability of Bruce 'B' RRS dual computer system

International Nuclear Information System (INIS)

Arsenault, J.E.; Manship, R.A.; Levan, D.G.

1995-07-01

The review presents an analysis of the Bruce 'B' Reactor Regulating System (RRS) Digital Control Computer (DCC) system, based on system documentation, significant event reports (SERs), question sets, and a site visit. The intent is to evaluate the reliability of the RRS DCC and to identify the possible scenarios that could lead to a serious process failure. The evaluation is based on three relatively independent analyses, which are integrated and presented in the form of Conclusions and Recommendations

Innovations in RCM at Bruce B

International Nuclear Information System (INIS)

Hill, E.S.; Doyle, E.K.

1996-01-01

The use of RCM techniques have begun to change maintenance practice at Bruce B. This paper discusses innovative practices begun recently. Bruce B has decided to evaluate plant systems using different methods based on the effects of system failure. This approach reduces costs, by using a streamlined method, while maintaining the accuracy of analysis. In addition, the approach increases the likelihood that program recommendations will be implemented by the maintenance department by providing maintenance craft with input to the process. Bruce B has also developed techniques to accelerate the analysis process by evaluating analyses performed at other units. These innovations have been successful piloted at the station

Future plans for performance analysis and maintenance/inspection optimization of shutoff rods based on the case study of Bruce Power Unit-3 Shutoff Rod 5 inspection

International Nuclear Information System (INIS)

Nasimi, E.; Gabbar, H.A.

2011-01-01

Shutdown System 1 (SDS1) is a preferred method for a quick shutdown of nuclear fission process in CANDU (CANada Deuterium Uranium) reactor units. Failure of a routine SDS1 safety test during Fall 2009 outage resulted in the need to develop and execute a new methodology for Shutoff Rod inspection and re-evaluate the known degradation mechanisms and failure modes. This paper describes the development of this methodology and the obtained results. It also proposes several alternative solutions for the future performance analysis and maintenance/inspection optimization for SDS1 Shutoff Rods based on the Bruce Power Unit-3 Shutoff Rod 5 case study. (author)

Bruce A refurbishment - an update

International Nuclear Information System (INIS)

Liddle, R.

2007-01-01

Running slightly ahead of schedule on the critical path work, the Bruce A Restart Project has not been without its challenges. About a dozen major contractors with a workforce of 1,700 tradespeople share space inside the Units 1 and 2 Construction Island. They share support services, provided by project management contractor AMEC NCL, and they share the consequences when one part of the project advances ahead of schedule or another falls behind. They also share Bruce Power's safety values and are well on their way to surpassing five-million hours without an acute lost time injury. (author)

Fuel deposits, chemistry and CANDU® reactor operation

International Nuclear Information System (INIS)

Roberts, J.G.

2014-01-01

'Hot conditioning' is a process which occurs as part of commissioning and initial start-up of each CANDU® reactor, the first being the Nuclear Power Demonstration - 2 reactor (NPD). Later, understanding of the cause of the failure of the Pickering Unit 1 G16 fuel channelled to a revised approach to 'hot conditioning', initially demonstrated on Bruce Unit 5. The difference being that during 'hot conditioning' of CANDU® heat transport systems fuel was not in-core until Bruce Unit 5. The 'hot conditioning' processes will be briefly described along with the consequences to fuel. (author)

Fuel handling solutions to power pulse at Bruce NGS A

International Nuclear Information System (INIS)

Day, R.C.

1996-01-01

In response to the discovery of the power pulse problem in March of 1993, Bruce A has installed flow straightening shield plugs in the inner zone channels of all units to partially reduce the gap and gain an increase in reactor power to 75%. After review and evaluation of solutions to manage the gap, including creep compensators and long fuel bundles, efforts have focused on a different solution involving reordering the fuel bundles to reverse the burnup profile. This configuration is maintained by fuelling with the flow and providing better support to the highly irradiated downstream fuel bundles by changing the design of the outlet shield plug. Engineering changes to the fuel handling control system and outlet shield plug are planned to be implemented starting in June 1996, thereby eliminating the power pulse problem and restrictions on reactor operating power. (author). 2 refs., 1 tab., 2 figs

Fuel handling solutions to power pulse at Bruce NGS A

Energy Technology Data Exchange (ETDEWEB)

Day, R C [Ontario Hydro, Tiverton, ON (Canada). Bruce Nuclear Generating Station-A

1997-12-31

In response to the discovery of the power pulse problem in March of 1993, Bruce A has installed flow straightening shield plugs in the inner zone channels of all units to partially reduce the gap and gain an increase in reactor power to 75%. After review and evaluation of solutions to manage the gap, including creep compensators and long fuel bundles, efforts have focused on a different solution involving reordering the fuel bundles to reverse the burnup profile. This configuration is maintained by fuelling with the flow and providing better support to the highly irradiated downstream fuel bundles by changing the design of the outlet shield plug. Engineering changes to the fuel handling control system and outlet shield plug are planned to be implemented starting in June 1996, thereby eliminating the power pulse problem and restrictions on reactor operating power. (author). 2 refs., 1 tab., 2 figs.

GRA model development at Bruce Power

International Nuclear Information System (INIS)

Parmar, R.; Ngo, K.; Cruchley, I.

2011-01-01

In 2007, Bruce Power undertook a project, in partnership with AMEC NSS Limited, to develop a Generation Risk Assessment (GRA) model for its Bruce B Nuclear Generating Station. The model is intended to be used as a decision-making tool in support of plant operations. Bruce Power has recognized the strategic importance of GRA in the plant decision-making process and is currently implementing a pilot GRA application. The objective of this paper is to present the scope of the GRA model development project, methodology employed, and the results and path forward for the model implementation at Bruce Power. The required work was split into three phases. Phase 1 involved development of GRA models for the twelve systems most important to electricity production. Ten systems were added to the model during each of the next two phases. The GRA model development process consists of developing system Failure Modes and Effects Analyses (FMEA) to identify the components critical to the plant reliability and determine their impact on electricity production. The FMEAs were then used to develop the logic for system fault tree (FT) GRA models. The models were solved and post-processed to provide model outputs to the plant staff in a user-friendly format. The outputs consisted of the ranking of components based on their production impact expressed in terms of lost megawatt hours (LMWH). Another key model output was the estimation of the predicted Forced Loss Rate (FLR). (author)

Water lancing of Bruce-A Unit 3 and 4 steam generators

International Nuclear Information System (INIS)

Puzzuoli, F.V.; Murchie, B.; Allen, S.

1995-01-01

During the Bruce-A 1993 Unit 4 and 1994 Unit 3 outages, three water lancing operations were carried out along with chemical cleaning as part of the station boiler refurbishment program. The water lancing activities focused on three boiler areas.. 1) support plates to clean partially or completely blocked broach holes and prevent boiler water level oscillations, 2) hot leg U-bend supports (HLUBS) to remove deposits contributing to boiler tube stress corrosion cracking (SCC) and 3) tube sheets to dislodge sludge piles that potentially threaten boiler tube integrity and to flush out post chemical cleaning insoluble residues. The combination of water lancing and chemical cleaning effectively reduced broach hole blockage from up to 100% to 0-10% or less. As a result, boilers in Units 3 and 4 will operate for some time to come without concerns over water level oscillations. However, deposits remained in most tube support plate land areas. (author)

AECB assessment of Bruce A comments on the proposed STPA tritium releases

International Nuclear Information System (INIS)

Gerdingh, R.F.

1995-04-01

In 1993, Ontario Hydro submitted a proposal to the AECB to discharge slightly tritiated water, arising from the steam supplied by Bruce A, through the Steam Transformer Plant A (STPA) discharge lines. The purpose of the proposal is to eliminate the current practice of shipping this water back to Bruce A where it is discharged through the active liquid waste discharge line. One of the STPA lines discharges to the Bruce A intake channel. A small fraction of this water is processed and used as drinking water at Bruce A. At our request, Bruce A management informed Bruce A personnel of the proposal and gave them the opportunity to raise concerns. As part of our evaluation, we assessed those concerns and concluded that they were not an impediment to accept this proposal. This report documents our assessment of the concerns expressed. (author)

AECB assessment of Bruce A comments on the proposed STPA tritium releases

Energy Technology Data Exchange (ETDEWEB)

Gerdingh, R F

1995-04-01

In 1993, Ontario Hydro submitted a proposal to the AECB to discharge slightly tritiated water, arising from the steam supplied by Bruce A, through the Steam Transformer Plant A (STPA) discharge lines. The purpose of the proposal is to eliminate the current practice of shipping this water back to Bruce A where it is discharged through the active liquid waste discharge line. One of the STPA lines discharges to the Bruce A intake channel. A small fraction of this water is processed and used as drinking water at Bruce A. At our request, Bruce A management informed Bruce A personnel of the proposal and gave them the opportunity to raise concerns. As part of our evaluation, we assessed those concerns and concluded that they were not an impediment to accept this proposal. This report documents our assessment of the concerns expressed. (author).

Instrument calibration optimization at Bruce Power: ECI loops

International Nuclear Information System (INIS)

Chugh, V.; Angelova, M.; Ghias, S.; Parmar, R.; Wang, V.; Xie, H.; Higgs, J.; Schut, J.; Cruchley, I.

2011-01-01

Most instruments in a nuclear power plant are calibrated at regular intervals to ensure consistency with the assumptions in the plant Technical Specifications and/or Safe Operating Envelope (SOE) compliance limits (e.g., As-Found Tolerance). In the Instrument Uncertainty Calculations (IUC), As-Found Tolerance for instrument drift is estimated based on statistical analysis of As-Found and As-Left calibration data such as that carried out for Bruce NGS by EPRI (Electric Power Research Institute) in 1998. Bruce specific drift values were found to compare favorably with industry benchmarks. Recently a significant amount of work has been done by EPRI and IAEA (International Atomic Energy Agency) on extending calibration intervals of safety related instruments. Reduction in calibration frequency reduces time commitments on the part of Authorized Nuclear Operators and safety system qualified Control Maintenance Technicians, and allows more schedule flexibility. To establish the proof of concept, As-Left/As-Found tolerances and available margins have been evaluated for the Bruce B Emergency Coolant Injection (ECI) system instrument loops to determine whether an extension of the calibration period from one or two year to three years is justifiable on the basis that these loops will still be in compliance with SOE. The analysis showed that 60% of instruments in the ECI system are qualified for calibration interval extension up to three years. Sensitivity assessment of the effect of proposed changes in calibration intervals for 60% of the instruments on the ECI system unavailability has also been performed using the current Bruce Power ECI unavailability model. The results show that, the largest ECI Predicted Future Unavailability (PFU) is 9.2E-4 year/year for in-core LOCA accident. This value is still below the target unavailability of 1.0E-3 year/year. (author)

Instrument calibration optimization at Bruce Power: ECI loops

Energy Technology Data Exchange (ETDEWEB)

Chugh, V.; Angelova, M.; Ghias, S.; Parmar, R.; Wang, V.; Xie, H. [AMEC NSS, Toronto, Ontario (Canada); Higgs, J.; Schut, J.; Cruchley, I. [Bruce Power, Tiverton, Ontario (Canada)

2011-07-01

Most instruments in a nuclear power plant are calibrated at regular intervals to ensure consistency with the assumptions in the plant Technical Specifications and/or Safe Operating Envelope (SOE) compliance limits (e.g., As-Found Tolerance). In the Instrument Uncertainty Calculations (IUC), As-Found Tolerance for instrument drift is estimated based on statistical analysis of As-Found and As-Left calibration data such as that carried out for Bruce NGS by EPRI (Electric Power Research Institute) in 1998. Bruce specific drift values were found to compare favorably with industry benchmarks. Recently a significant amount of work has been done by EPRI and IAEA (International Atomic Energy Agency) on extending calibration intervals of safety related instruments. Reduction in calibration frequency reduces time commitments on the part of Authorized Nuclear Operators and safety system qualified Control Maintenance Technicians, and allows more schedule flexibility. To establish the proof of concept, As-Left/As-Found tolerances and available margins have been evaluated for the Bruce B Emergency Coolant Injection (ECI) system instrument loops to determine whether an extension of the calibration period from one or two year to three years is justifiable on the basis that these loops will still be in compliance with SOE. The analysis showed that 60% of instruments in the ECI system are qualified for calibration interval extension up to three years. Sensitivity assessment of the effect of proposed changes in calibration intervals for 60% of the instruments on the ECI system unavailability has also been performed using the current Bruce Power ECI unavailability model. The results show that, the largest ECI Predicted Future Unavailability (PFU) is 9.2E-4 year/year for in-core LOCA accident. This value is still below the target unavailability of 1.0E-3 year/year. (author)

An Appreciation of the Scientific Life and Acheivements of Bruce Merrifield

Energy Technology Data Exchange (ETDEWEB)

Mitchell, R

2007-06-15

Bruce Merrifield's scientific biography, 'Life During a Golden Age of Peptide Chemistry: The Concept and Development of Solid-Phase Peptide Synthesis', provides a history of solid phase-peptide synthesis (SPPS) from 1959 to 1993 [1]. While many readers will be familiar with SPPS literature after 1963, the inclusion of unpublished material from Merrifield's early laboratory notebooks opens a fascinating window on the development of SPPS from the formulation of concept in 1959 (p. 56, ref. 1) to the synthesis of a tetrapeptide four years later [2]. This early period was characterized by slow progress interrupted by numerous setbacks that led Bruce to later record (p. 90, ref. 1): 'At the end of the first two years the results were so poor, I wonder what made me think that this approach would ever succeed; but from the outset I had a strong conviction that this was a good idea, and I am glad that I stayed with it long enough'. Garland Marshall, Bruce's first graduate student (1963-1966), as well as later colleagues, were essentially unaware of the many highways, byways and dead ends that Bruce had explored in the early years [3].

Fans af Bruce

DEFF Research Database (Denmark)

Vaaben, Nana Katrine

2007-01-01

Analysen viser, hvordan det samme ritual under en koncert forener og opdeler de fans, der orienterer sig mod Bruce Springsteen. På den ene side forener ritualet hele publikum i en stor fælles "Intimitet for mange" og på den anden side splitter det dem, fordi det bliver tydeligt, hvem der er de...... rigtige fans, og hvem der tilhører "pøbelen"....

Nuclear process steam for industry: potential for the development of an Industrial Energy Park adjacent to the Bruce Nuclear Power Development

Energy Technology Data Exchange (ETDEWEB)

Seddon, W A

1981-11-01

This report summarizes the results of an industrial survey jointly funded by the Bruce County Council, the Ontario Energy Corporation, Atomic Energy of Canada Limited and conducted with the cooperation of Ontario Hydro and the Ontario Ministry of Industry and Tourism. The objective of the study was to identify and assess the future needs and interest of energy-intensive industries in the concept of an Industrial Energy Park adjacent tof the Bruce Nuclear Power Development. The proposed Energy Park would capitalize on the infrastructure of the existing CANDU reactors and Ontario Hydro's proven and unique capability to produce steam, as well as electricity, at a cost currently about half that from a comparable coal-fired station.

Bruce A refurbishment - preparatory work completed, major tasks to begin soon

International Nuclear Information System (INIS)

Boyd, F.

2006-01-01

Over the past year Bruce Power has been planning and organizing for an extensive refurbishment of the Units 1 and 2 of the Bruce A station. Now the company and its several major contractors are ready to proceed with the most challenging aspects of the actual work. The largest tasks are the replacement of the 8 steam generators and of the 480 complete fuel channels in each unit Bruce Power has created a separate website connected to their basic one to provide ongoing information about the progress of the work. The following brief note is intended to provide an outline of this challenging refurbishment program and to invite readers to visit this website to follow its progress. To provide background the writer was accorded an informative and interesting tour of the units by Rob Liddle, of Bruce Power, on September 28, 2006 the day after the ceremony commemorating the Douglas Point station held at the Bruce site. (author)

Fuel deposits, chemistry and CANDU reactor operation

International Nuclear Information System (INIS)

Roberts, J.G.

2013-01-01

'Hot conditioning' is a process which occurs as part of commissioning and initial start-up of each CANDU reactor, the first being the Nuclear Power Demonstration-2 reactor (NPD). Later, understanding of the cause of the failure of the Pickering Unit 1 G16 fuel channel led to a revised approach to 'hot conditioning', initially demonstrated on Bruce Unit 5, and subsequently utilized for each CANDU unit since. The difference being that during 'hot conditioning' of CANDU heat transport systems fuel was not in-core until Bruce Unit 5. The 'hot conditioning' processes will be briefly described along with the consequences to fuel. (author)

Steam generator tubesheet waterlancing at Bruce B

Energy Technology Data Exchange (ETDEWEB)

Persad, R. [Babcock and Wilcox Canada, Cambridge, Ontario (Canada); Eybergen, D. [Bruce Power, Tiverton, Ontario (Canada)

2006-07-01

High pressure water cleaning of steam generator secondary side tubesheet surfaces is an important and effective strategy for reducing or eliminating under-deposit chemical attack of the tubing. At the Bruce B station, reaching the interior of the tube bundle with a high-pressure water lance is particularly challenging due to the requirement to setup on-boiler equipment within the containment bellows. This paper presents how these and other design constraints were solved with new equipment. Also discussed is the application of new high-resolution inter-tube video probe capability to the Bruce B steam generator tubesheets. (author)

Fuel bundle to pressure tube fretting in Bruce and Darlington

Energy Technology Data Exchange (ETDEWEB)

Norsworthy, A G; Ditschun, A [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1996-12-31

As the fuel channel elongates due to creep, the fuel string moves relative to the inlet until the fuel pads at the inboard end eventually separate from the spacer sleeve, and the fuel resides on the burnish mark of the pressure tube. The bundle is then supported in a fashion which contributes to increased levels of vibration. Those pads which (due to geometric variation) have contact loads with the pressure tube within a certain range, vibrate, and cause significant fretting on the burnish mark, and further along at the midplane of the bundle. Inspection of the pressure tubes in Bruce A, Bruce B, and Darlington has revealed fret damage up to 0.55 mm at the burnish mark and slightly lower than this at the inlet bundle midplane. To date, all fret marks have been dealt with successfully without the need for tube replacement, but a program of work has been initiated to understand the mechanism and reduce the fretting. Such understanding is necessary to guide future design changes to the fuel bundle, to guide future inspection programs, to guide maintenance programs, and for longer term strategic planning. This paper discusses how the understanding of fretting has evolved and outlines a current hypothesis for the mechanism of fretting. The role of bundle geometry, excitation forces, and reactor conditions are reviewed, along with options under consideration to mitigate damage. (author). 4 refs., 2 tabs., 13 figs.

Fuel bundle to pressure tube fretting in Bruce and Darlington

International Nuclear Information System (INIS)

Norsworthy, A.G.; Ditschun, A.

1995-01-01

As the fuel channel elongates due to creep, the fuel string moves relative to the inlet until the fuel pads at the inboard end eventually separate from the spacer sleeve, and the fuel resides on the burnish mark of the pressure tube. The bundle is then supported in a fashion which contributes to increased levels of vibration. Those pads which (due to geometric variation) have contact loads with the pressure tube within a certain range, vibrate, and cause significant fretting on the burnish mark, and further along at the midplane of the bundle. Inspection of the pressure tubes in Bruce A, Bruce B, and Darlington has revealed fret damage up to 0.55 mm at the burnish mark and slightly lower than this at the inlet bundle midplane. To date, all fret marks have been dealt with successfully without the need for tube replacement, but a program of work has been initiated to understand the mechanism and reduce the fretting. Such understanding is necessary to guide future design changes to the fuel bundle, to guide future inspection programs, to guide maintenance programs, and for longer term strategic planning. This paper discusses how the understanding of fretting has evolved and outlines a current hypothesis for the mechanism of fretting. The role of bundle geometry, excitation forces, and reactor conditions are reviewed, along with options under consideration to mitigate damage. (author). 4 refs., 2 tabs., 13 figs

David Bruce Payton : väikeriigid mõistavad üksteist / David Bruce Payton ; interv. Marianne Mikko

Index Scriptorium Estoniae

Payton, David Bruce

2003-01-01

Uus-Meremaa suursaadik Eestis David Bruce Payton talupidaja toetamise loobumisest Uus-Meremaal, Uus-Meremaa põllumajandussektorist, veinidest, ekspordist, Eesti saamisest EL-i ja NATO liikmeks, Uus-Meremaa rahvastikust, elatustasemest, Iraagi võimalikust ründamisest, Põhja-Koreast

Bruce Springsteen as a Symbol

DEFF Research Database (Denmark)

Gitz-Johansen, Thomas

2018-01-01

The article explores how Bruce Springsteen and his music function as a symbol. The article first presents the Jungian theory of symbols and of music as symbol. The central argument of the article is that, by functioning symbolically, Springsteen has the potential to influence the psyche of his au...

Characterization and dissolution studies of Bruce Unit 3 steam generator secondary side deposits

International Nuclear Information System (INIS)

Semmler, J.

1998-01-01

The physical and chemical properties of secondary side steam generator deposits in the form of powder and flake obtained from Bruce Nuclear Generating Station A (BNGS A) Unit 3 were studied. The chemical phases present in both types of deposits, collected prior to the 1994 chemical cleaning during the pre-clean water lancing campaign, were magnetite (Fe 3 O 4 ), metallic copper (Cu), hematite (Fe 2 O 3 ) and cuprous oxide (Cu 2 O). The major difference between the chemical composition of the powder and the flake was the presence of zinc silicate (Zn 2 SiO 4 ) and several unidentified silicate phases containing Ca, Al, Mn, and Mg in the flake. The flake deposit had high hardness values, high electrical resistivity, low porosity and a lower dissolution rate in the EPRI-SGOG (Electric Power Research Institute-Steam Generator Owner's Group) chemical cleaning solvents compared to the powder deposit. Differences in the deposit properties after chemical cleaning of the Unit 3 steam generators and after laboratory cleaning were noted. The presence of silicates in the deposit inhibit magnetite dissolution

Safety benefits from CANDU reactor replacement - a case study

International Nuclear Information System (INIS)

Mottram, R.; Millard, J.W.F.; Purdy, P.

2011-01-01

Both total core replacement and core retubing have been used in the CANDU industry. For future plant refurbishments, based on experience both in new construction and in recent refurbishments, the concept of total core replacement has been revisited. This builds on practices for replacement of other large plant equipment like boilers. The Bruce CANDU reactors, with their local shield tanks built around the Calandria and containment closely located around that Calandria Shield Tank Assembly (CSTA), are believed to be good candidates for core replacement. A structured process was used to design a replacement CSTA suitable for Bruce A use. The work started with a study of opportunities for safety enhancements in the core. This progressed into design studies and related design assist safety analysis on the reactor. A key element of the work involved consideration of how verified features from later CANDU designs, and from our new reactor design work, could be tailored to fit this replacement core. The replacement reactor core brings in structural improvements in both calandria and end shield, and safety improvements like the natural circulation enhancing moderator cooling layout and further optimized reactivity layouts to improve shutdown system performance. Bruce Power are currently studying the business implications of this and retube techniques as part of preparation for future refurbishments. The work explained in this paper is in the context of the safety related changes and the work to choose and quantify them. (author)

Safety benefits from CANDU reactor replacement. A case study

International Nuclear Information System (INIS)

Mottram, R.; Millard, J.W.F.; Purdy, P.

2011-01-01

Both total core replacement and core retubing have been used in the CANDU industry. For future plant refurbishments, based on experience both in new construction and in recent refurbishments, the concept of total core replacement has been revisited. This builds on practices for replacement of other large plant equipment like boilers. The Bruce CANDU reactors, with their local shield tanks built around the Calandria and containment closely located around that Calandria Shield Tank Assembly (CSTA), are believed to be good candidates for core replacement. A structured process was used to design a replacement CSTA suitable for Bruce A use. The work started with a study of opportunities for safety enhancements in the core. This progressed into design studies and related design assist safety analysis on the reactor. A key element of the work involved consideration of how verified features from later CANDU designs, and from our new reactor design work, could be tailored to fit this replacement core. The replacement reactor core brings in structural improvements in both calandria and end shield, and safety improvements like the natural circulation enhancing moderator cooling layout and further optimized reactivity layouts to improve shutdown system performance. Bruce Power are currently studying the business implications of this and retube techniques as part of preparation for future refurbishments. The work explained in this paper is in the context of the safety related changes and the work to choose and quantify them. (author)

Universal delivery machine - design of the Bruce and Darlington heads

International Nuclear Information System (INIS)

Gray, M.G.; Brown, R.

2003-01-01

The Universal Delivery Machine (UDM) was designed and supplied to reduce the time required to perform channel inspection services. The Bruce UDM was the first to be completed followed by Pickering and Darlington. The Bruce and Darlington machines are nearly identical. Design concepts applied include a rotating, multiple tool station magazine, a rigid chain driving telescoping rams, a common drive package, and an external support frame to meet seismic qualification requirements. (author)

Bruce used fuel dry storage project evolution from Pickering to Bruce

International Nuclear Information System (INIS)

Young, R.E.

1996-01-01

Additional fuel storage capacity is required at Bruce Nuclear Generating Station, which otherwise would soon fill up all its pool storage capacity. The recommended option was to use a dry storage container similar to that at Pickering. The changes made to the Pickering type of container included: fuel to be stored in trays; the container's capacity increased to 600 bundles; the container's lid to be changed to a metal one; the single concrete lid to be changed to a double metal lid system; the container not to be transportable; the container would be dry-loaded. 7 figs

Bruce used fuel dry storage project evolution from Pickering to Bruce

Energy Technology Data Exchange (ETDEWEB)

Young, R E [Ontario Hydro, Tiverton, ON (Canada). Bruce Nuclear Generating Station-A

1997-12-31

Additional fuel storage capacity is required at Bruce Nuclear Generating Station, which otherwise would soon fill up all its pool storage capacity. The recommended option was to use a dry storage container similar to that at Pickering. The changes made to the Pickering type of container included: fuel to be stored in trays; the container`s capacity increased to 600 bundles; the container`s lid to be changed to a metal one; the single concrete lid to be changed to a double metal lid system; the container not to be transportable; the container would be dry-loaded. 7 figs.

Bruce A - performance power

Energy Technology Data Exchange (ETDEWEB)

Boucher, P. [Bruce Power, Tiverton, ON (Canada)

2015-07-01

This paper discusses the strategy for improving performance at Bruce Power. The key to excellence is changing behaviours. Reinforcing and enforcing expectations, aligned with the 2015 operating to the Highest Standards Site Initiative. Long term equipment strategies, supported by the 2015 Equipment Health Site Initiative, individual and group accountability for online/outage Work Management, with further gains through 2015 Maintenance Alignment and Resource Strategy (MARS) Site Initiative. Results showed human performance improvement, more reliable and predictable units and outage performance improvement.

The Bruce Medalists

Science.gov (United States)

Tenn, J. S.

2001-12-01

The Astronomical Society of the Pacific (ASP) has presented the Catherine Wolfe Bruce gold medal for lifetime contributions to astronomy most years since 1898. The 94 medalists include most of the scientists whose work has greatly changed astronomy since the late nineteenth century: Huggins, Pickering, Campbell, Hale, Eddington, Russell, Adams, Slipher, Hertzsprung, Hubble, Shapley, Oort, Baade, ... Major exceptions include those who died young, those who worked in teams, and, in the early years, women. Mathematicians appear to have been as likely to be honored as astronomers from the beginning, but the fortunes of physicist nominees have varied. The nomination process is an unusual one, with the directors of six observatories, three in the U.S. and three abroad, asked to nominate up to three candidates each year. For the first six decades the observatories rarely varied, and directors had long tenures. They nominated the same individuals repeatedly. Now both observatories and their directors vary regularly. Much can be learned about the changes in astronomy from the late nineteenth century, when observers worked alone with long refractors and a theorist could spend a lifetime computing the orbit of one comet, to the present, when most papers have multiple authors and a single project may include millions of objects. For example, celestial mechanics was the specialty of many of the early medalists but none since 1966. I have posted photographs, brief biographies, extensive bibliographies, and links to publications by and about all of the medalists, from Simon Newcomb in 1898 to Hans Bethe in 2001, at http://phys-astro.sonoma.edu/BruceMedalists/. I will discuss a bit of the history of the medal and some of the medalists.

An intelligent safety system concept for future CANDU reactors

International Nuclear Information System (INIS)

Hinds, H.W.

1980-01-01

A review of the current Regional Over-power Trip (ROPT) system employed on the Bruce NGS-A reactors confirmed the belief that future reactors should have an improved ROPT system. We are developing such an 'intelligent' safety system. It uses more of the available information on reactor status and employs modern computer technology. Fast triplicated safety computers compute maps of fuel channel power, based on readings from prompt-responding flux detectors. The coefficients for this calculation are downloaded periodically from a fourth supervisor computer. These coefficients are based on a detailed 3-D flux shape derived from physics data and other plant information. A demonstration of one of three safety channels of such a system is planned. (auth)

Non intrusive check valve diagnostics at Bruce A

International Nuclear Information System (INIS)

Marsch, S.P.

1997-01-01

Bruce A purchased non intrusive check valve diagnostic equipment in 1995 to ensure operability and availability of critical check valves in the Station. Diagnostics can be used to locate and monitor check valve degradation modes. Bruce A initiated a pilot program targeting check valves with flow through them and ones that completed open or close cycles. Approaches to determine how to confirm operability of passive check valves using non intrusive techniques were explored. A sample population of seventy-three check valves was selected to run the pilot program on prior to complete implementation. The pilot program produced some significant results and some inconclusive results. The program revealed a major finding that check valve performance modeling is required to ensure continuous operability of check valves. (author)

Non intrusive check valve diagnostics at Bruce A

Energy Technology Data Exchange (ETDEWEB)

Marsch, S.P. [Ontario Hydro, Bruce Nuclear Generating Station A, Tiverton, ON (Canada)

1997-07-01

Bruce A purchased non intrusive check valve diagnostic equipment in 1995 to ensure operability and availability of critical check valves in the Station. Diagnostics can be used to locate and monitor check valve degradation modes. Bruce A initiated a pilot program targeting check valves with flow through them and ones that completed open or close cycles. Approaches to determine how to confirm operability of passive check valves using non intrusive techniques were explored. A sample population of seventy-three check valves was selected to run the pilot program on prior to complete implementation. The pilot program produced some significant results and some inconclusive results. The program revealed a major finding that check valve performance modeling is required to ensure continuous operability of check valves. (author)

Fuelling with flow at Bruce A

Energy Technology Data Exchange (ETDEWEB)

Gray, M G [Canadian General Electric Co. Ltd., Peterborough, ON (Canada)

1997-12-31

Fuelling with flow is the solution chosen by Bruce A to overcome the potential power pulse caused by a major inlet header failure. Fuelling with flow solves the problem by rearranging the core to place new fuel at the channel inlet and irradiated fuel at the channel outlet. The change has a significant impact on the Bruce A fuel handling system which was designed primarily to do on power fuelling in the against flow direction. Mechanical changes to the fuelling machine include a modification to the existing ram head and the replacement of standard fuel carriers with new fuelling with flow fuel carriers having the capability of opening the channel latch. Changes to the control system are more involved. A new set of operational sequences are required for both the upstream and downstream fuelling machines to achieve the fuel change. Steps based on sensitive ram push are added to reduce the risk of failing to close the latch at the correct position to properly support the fuel string. Changes are also required to the protective interlocks to allow fuelling with flow and reduce risk. A new fuel string supporting shield plug was designed and tested to reduce the risk of endplate cracking that could occur on the irradiated bundle that would have been supported directly by the channel latch. Some operational changes have been incorporated to accommodate this new shield plug. Considerable testing has been carried out on all aspects of fuel handling where fuelling with flow differs from the reference fuelling against flow. (author). 3 figs.

Bruce NGS A Unit 4 preheater divider plate failure

International Nuclear Information System (INIS)

Landridge, M.; McInnes, D.

1995-01-01

On May 19, 1995, without any prior operational indications, Bruce A discovered preheater divider plate damage in Unit 4 that had the potential to have a major impact on the continued safe operation of the station. Further investigations indicated that Unit 4 may have been operating with this damage for as long as ten years. In the two months following the discovery, Bruce A has procured and replaced the 4 divider plates, located most of the missing pieces, retrieved pieces from the PHT system, investigated historical operational information, performed detailed analytical investigations, investigated root cause, performed in-situ and mock-up testing, updated operational procedures and installed DP monitoring equipment

Measurements of Sheath Temperature Profiles in Bruce LVRF Bundles Under Post-Dryout Heat Transfer Conditions in Freon

International Nuclear Information System (INIS)

Guo, Y.; Bullock, D.E.; Pioro, I.L.; Martin, J.

2006-01-01

An experimental program has been completed to study the behaviour of sheath wall temperatures in the Bruce Power Station Low Void Reactivity Fuel (shortened hereafter to Bruce LVRF) bundles under post-dryout (PDO) heat-transfer conditions. The experiment was conducted with an electrically heated simulator of a string of nine Bruce LVRF bundles, installed in the MR-3 Freon heat transfer loop at the Chalk River Laboratories (CRL), Atomic Energy of Canada Limited (AECL). The loop used Freon R-134a as a coolant to simulate typical flow conditions in CANDU R nuclear power stations. The simulator had an axially uniform heat flux profile. Two radial heat flux profiles were tested: a fresh Bruce LVRF profile and a fresh natural uranium (NU) profile. For a given set of flow conditions, the channel power was set above the critical power to achieve dryout, while heater-element wall temperatures were recorded at various overpower levels using sliding thermocouples. The maximum experimental overpower achieved was 64%. For the conditions tested, the results showed that initial dryout occurred at an inner-ring element at low flows and an outer-ring element facing internal subchannels at high flows. Dry-patches (regions of dryout) spread with increasing channel power; maximum wall temperatures were observed at the downstream end of the simulator, and immediately upstream of the mid-bundle spacer plane. In general, maximum wall temperatures were observed at the outer-ring elements facing the internal subchannels. The maximum water-equivalent temperature obtained in the test, at an overpower level of 64%, was significantly below the acceptable maximum temperature, indicating that the integrity of the Bruce LVRF will be maintained at PDO conditions. Therefore, the Bruce LVRF exhibits good PDO heat transfer performance. (authors)

AECB staff annual assessment of the Bruce A Nuclear Generating Station for the year 1995

International Nuclear Information System (INIS)

1996-06-01

The Atomic Energy Control Board conducts a staff assessment of safety at Bruce Nuclear Generating Station A for 1995. On-site Project Officers and Ottawa based specialists monitored the station throughout the year. Ontario Hydro operated Bruce A safely in 1995, maintaining the risk to workers and the public at an acceptably low level. Radiation doses to workers and releases to the environment were well below regulatory limits. However, Ontario Hydro must improve contamination control at Bruce A. Special safety system performance a Bruce A was less than adequate. The negative pressure containment system and units 4's shutdown system two exceeded unavailability targets in 1995. However, we are satisfied Ontario Hydro is taking appropriate action to correct this. 5 tabs., 5 figs

Bruce Power's nuclear pressure boundary quality assurance program requirements, implementation and transition

International Nuclear Information System (INIS)

Krane, J.C.

2009-01-01

The development of a full scope nuclear pressure boundary quality assurance program in Canada requires extensive knowledge of the structure and detailed requirements of codes and standards published by the Canadian Standards Association (CSA) and American Society of Mechanical Engineers (ASME). Incorporation into company governance documents and implementation of these requirements while managing the transition to more recent revisions of these codes and standards represents a significant challenge for Bruce Power, Canada's largest independent nuclear operator. This paper explores the key developments and innovative changes that are used to ensure successful regulatory compliance and effective implementation of the Bruce Power Pressure Boundary Quality Assurance Program. Challenges and mitigating strategies to sustain this large compliance based program at Bruce Power's 8 unit nuclear power plant site will also be detailed. (author)

Structural and metamorphic evolution of the Mid-Late Proterozoic Rayner Complex, Cape Bruce, East Antarctica

International Nuclear Information System (INIS)

Dunkley, D.J.; Clarke, G.L.; White, R.W.

2002-01-01

Granulite to transitional granulite facies gneisses exposed at Cape Bruce, Rayner Complex, East Antarctica, record three main orogenic/magmatic phases: (1) intrusion of c. 1000-980 Ma felsic orthogneisses into Mid-Proterozoic metasediments, contemporary with the development of north-trending reclined to recumbent folds; (2) extensive c. 980-900 Ma felsic magmatism, including equivalents of the Mawson Charnockite, which accompanied the development of upright, east-northeast-trending folds; and (3) ultramylonite zones of uncertain age. The first two phases are known as the Rayner Structrual Episode, the effects of which are similar in rocks to the east of Cape Bruce, at Mawson, and in the northern Prince Charles Mountains. Archaean rocks immediately to the west of Cape Bruce were tectonically reworked during the Rayner Structural Episode. The first orogenic phase is inferred to represent the collision between a wedge-shaped Proterozoic block comprising rocks of the Mawson Coast and Eastern Ghats Province, with the Archaean Napier Complex. The second orogenic phase included a major period of crustal growth through emplacement of the Mawson Charnockite and equivalents. (author). 41 refs., 6 figs., 1 tab

Exercise testing of pre-school children using the Bruce treadmill protocol: new reference values

NARCIS (Netherlands)

M.H.M. van der Cammen-van Zijp (Monique); H. IJsselstijn (Hanneke); T. Takken (Tim); S.P. Willemsen (Sten); D. Tibboel (Dick); H.J. Stam (Henk); H.J.G. van den Berg-Emons (Rita)

2010-01-01

textabstractThe Bruce treadmill protocol is an often-used exercise test for children and adults. Few and mainly old normative data are available for young children. In this cross-sectional observational study we determined new reference values for the original Bruce protocol in children aged 4 and 5

Industrial process heat from CANDU reactors

International Nuclear Information System (INIS)

Hilborn, J.S.; Seddon, W.A.; Barnstaple, A.G.

1980-08-01

It has been demonstrated on a large scale that CANDU reactors can produce industrial process steam as well as electricity, reliably and economically. The advantages of cogeneration have led to the concept of an Industrial Energy Park adjacent to the Bruce Nuclear Power Development in the province of Ontario. For steam demands between 300,000 and 500,00 lb/h (38-63 kg/s) and an annual load factor of 80%, the estimated cost of nuclear steam at the Bruce site boundary is $3.21/MBtu ($3.04GJ), which is at least 30% cheaper than oil-fired steam at the same site. The most promising near term application of nuclear heat is likely to be found within the energy-intensive chemical industry. Nuclear energy can substitute for imported oil and coal in the eastern provinces if the price remains competitive, but low cost coal and gas in the western provinces may induce energy-intensive industries to locate near those sources of energy. In the long term it may be feasible to use nuclear heat for the mining and extraction of oil from the Alberta tar sands. (auth)

Geoscientific Characterization of the Bruce Site, Tiverton, Ontario

Science.gov (United States)

Raven, K.; Jackson, R.; Avis, J.; Clark, I.; Jensen, M.

2009-05-01

Ontario Power Generation is proposing a Deep Geologic Repository (DGR) for the long-term management of its Low and Intermediate Level Radioactive Waste (L&ILW) within a Paleozoic-age sedimentary sequence beneath the Bruce site near Tiverton, Ontario, Canada. The concept envisions that the DGR would be excavated at a depth of approximately 680 m within the Ordovician Cobourg Formation, a massive, dense, low- permeability, argillaceous limestone. Characterization of the Bruce site for waste disposal is being conducted in accordance with a four year multi-phase Geoscientific Site Characterization Plan (GSCP). The GSCP, initially developed in 2006 and later revised in 2008 to account for acquired site knowledge based on successful completion of Phase I investigations, describes the tools and methods selected for geological, hydrogeological and geomechanical site characterization. The GSCP was developed, in part, on an assessment of geoscience data needs and collection methods, review of the results of detailed geoscientific studies completed in the same bedrock formations found off the Bruce site, and recent international experience in geoscientific characterization of similar sedimentary rocks for long-term radioactive waste management purposes. Field and laboratory work related to Phase 1 and Phase 2A are nearing completion and have focused on the drilling, testing and monitoring of four continuously cored vertical boreholes through Devonian, Silurian, Ordovician and Cambrian bedrock to depths of about 860 mBGS. Work in 2009 will focus on drilling and testing of inclined boreholes to assess presence of vertical structure. The available geological, hydrogeological and hydrogeochemical data indicate the presence of remarkably uniform and predictable geology, physical hydrogeologic and geochemical properties over well separation distances exceeding 1 km. The current data set including 2-D seismic reflection surveys, field and lab hydraulic testing, lab petrophysical and

Waste repository planned for Bruce Site

International Nuclear Information System (INIS)

King, F.

2004-01-01

Ontario Power Generation (OPG) and Kincardine, the municipality nearest the Bruce site, have agreed in principal to the construction of a deep geologic repository for low and medium level radioactive waste on the site. The two parties signed the 'Kincardine Hosting Agreement' on October 13, 2004 to proceed with planning, seek regulatory approval and further public consultation of the proposed project. A construction Licence is not expected before 2013. (author)

Corrosion-product inventory: the Bruce-B secondary system

International Nuclear Information System (INIS)

Sawicki, J.A.; Price, J.; Brett, M.E.

1995-01-01

Corrosion inspection and corrosion-product characterization in water and steam systems are important for component and systems maintenance in nuclear power stations. Corrosion products are produced, released and redeposited at various sites in the secondary system. Depending on the alloys used in the condenser and feedwater heaters, particulate iron oxides and hydroxides can account for about 95-99% of the total corrosion-product transport. Where brass or cupro-nickel alloys are present, copper and zinc contribute significantly to the total transport and deposition. Particulates are transported by the feedwater to the steam generators, where they accumulate and can cause a variety of problems, such as loss of heat transfer capability through deposition on boiler tubes, blockage of flow through boiler-tube support plates and accelerated corrosion in crevices, either in deep sludge piles or at blocked tube supports. The influx of oxidized corrosion products may have a particularly adverse effect on the redox environment of steam generator tubing, thereby increasing the probability of localized corrosion and other degradation mechanisms. In this paper, there is a description of a survey of general corrosion deposits in Bruce-B, Units 5-8, which helps to identify the origin, evolution and inventory of corrosion products along the secondary system of Candu reactors

Status of the reliability centered maintenance program at Ontario Hydro's Bruce 'A' Nuclear Division

International Nuclear Information System (INIS)

Khan, I.

1995-01-01

Bruce A started a preventive maintenance (PM) quality improvement program in August of 1991. This initiative was taken to address the concerns expressed by the AECB and the Peer Audits finding. The concerns were on the quality of the Bruce A PM Program and its execution in the field. Reliability Centered Maintenance (RCM) analysis was selected as the PM program quality improvement and optimization technique. Therefore, RCM became a key component of Bruce A's Integrated PM program and maintenance strategy. As a result of RCM implementation, and improvements in the work planning and scheduling process, Bruce A is seeing downward trends in the corrective maintenance work load, maintenance preventable forced outages, overdue/missed PM tasks and corrective maintenance backlog. Control Room Operators have reported observing an improvement in systems and equipment response to transients. Other benefits include a documented, controlled and traceable PM program. In addition, the team approach required by RCM has started to improve staff confidence in the PM program which, in turn, is improving the compliance with the PM program. (author)

Evolution of CANDU reactor design

International Nuclear Information System (INIS)

Pon, G.A.

1978-08-01

The CANDU (CANada Deuterium Uranium) design had its begin-ings in the early 1950's with the preliminary engineering studies that led to the 20 MW(e) NPD (Nuclear Power Demonstration) and the 200 MW(e) Douglas Point station . The next decade saw the first operation of both these stations and the commitment of the 2000 MW(e) Pickering and 3000 MW(e) Bruce plants. The present decade has witnessed the excellent performance of Pickering and Bruce and commitments to construct Gentilly-2, Cordoba, Pt. Lepreau, Wolsung, Pickering B, Bruce B and Darlington. In most cases, successive CANDU designs have meant an increase in plant output. Evolutionary developments have been made to fit the requirements of higher ratings and sizes, new regulations, better reliability and maintainability and lower costs. These changes, which are described system by system, have been introduced in the course of engineering parallel reactor projects with overlapping construction schedules -circumstances which ensure close contact with the practical realities of economics, manufacturing functions, construction activities and performance in commissioning. Features for one project furnished alternative concepts for others still on the drawing board and the experience gained in the first application yielded a sound basis for its re-use in succeeding projects. Thus the experiences gained in NPD, Douglas Point, Gentilly-1 and KANUPP have contributed to Pickering and Bruce, which in turn have contributed to the design of Gentilly-2. (author)

Regional and site geological frameworks : proposed Deep Geologic Repository, Bruce County, Ontario

Energy Technology Data Exchange (ETDEWEB)

Raven, K.; Sterling, S.; Gaines, S.; Wigston, A. [Intera Engineering Ltd., Ottawa, ON (Canada); Frizzell, R. [Nuclear Waste Management Organization, Toronto, ON (Canada)

2009-07-01

The Nuclear Waste Management Organization is conducting geoscientific studies on behalf of Ontario Power Generation into the proposed development of a Deep Geologic Repository (DGR) for low and intermediate level radioactive waste (L and ILW) at the Bruce site, near Tiverton, Ontario. This paper presented a regional geological framework for the site that was based on a review of regional drilling; structural geology; paleozoic stratigraphy and sedimentology; a 3D geological framework model; a DGR geological site characterization model; bedrock stratigraphy and marker beds; natural fracture frequency data; and formation predictability. The studies have shown that the depth, thickness, orientation and rock quality of the 34 rock formations, members or units that comprise the 840 m thick Paleozoic bedrock sequence at the Bruce site are very uniform and predictable over distances of several kilometres. The proposed DGR will be constructed as an engineered facility comprising a series of underground emplacement rooms at a depth of 680 metres below ground within argillaceous limestones. The geoscientific studies are meant to provide a basis for the development of descriptive geological, hydrogeological and geomechanical models of the DGR site that will facilitate environmental and safety assessments. 11 refs., 3 tabs., 9 figs.

Fire fighting capability assessment program Bruce B NGS

International Nuclear Information System (INIS)

1995-05-01

This is a report on the completion of work relating to the assessment of the capability of Bruce B NGS to cope with a large fire incident. This included an evaluation of an exercise scenario that would simulate a large fire incident and of their fire plans and procedures. Finally the execution of fire plans by Bruce B NGS, as demonstrated by their application of human and material resources during a simulated large fire, was observed. The fire fighting equipment and the personal protective clothing and associated equipment that was in use was all of good quality and in good condition. There had also been notable improvement in communications equipment. Similarly, the human resources that had been assigned to fire fighting and rescue crews and that were available were more than adequate. Use of a logical incident command system, and the adoption of proper policy and procedures for radio communications were equally significant improvements. Practice should correct the breakdowns that occurred in these areas during the exercise. As well, there remains a need for the development of policy on fire fighting and rescue operations with more depth and clarity. In summary, the key point to be recognized is the degree of improvement that has been realized since the previous evaluation in 1990. Clearly the Emergency Response Teams organization of Bruce B NGS is evolving into an effective fire fighting force. Providing that the deficiencies identified in this report are addressed satisfactorily, Fire Cross is confident that the organization will have the capability to provide rescue and fire fighting services that will satisfy the need. 2 figs

Methyl iodide trapping efficiency of aged charcoal samples from Bruce-A emergency filtered air discharge systems

International Nuclear Information System (INIS)

Wren, J.C.; Moore, C.J.; Rasmussenn, M.T.; Weaver, K.R.

1999-01-01

Charcoal filters are installed in the emergency filtered air discharge system (EFADS) of multiunit stations to control the release of airborne radioiodine in the event of a reactor accident. These filters use highly activated charcoal impregnated with triethylenediamine (TEDA). The TEDA-impregnated charcoal is highly efficient in removing radioiodine from flowing airstreams. The iodine-removal efficiency of the charcoal is presumed to deteriorate slowly with age, but current knowledge of this effect is insufficient to predict with confidence the performance of aged charcoal following an accident. Experiments were performed to determine the methyl iodide removal efficiency of aged charcoal samples taken from the EFADS of Ontario Hydro's Bruce-A nuclear generating station. The charcoal had been in service for ∼4 yr. The adsorption rate constant and capacity were measured under post-loss-of-coolant accident conditions to determine the efficiency of the aged charcoal. The adsorption rate constants of the aged charcoal samples were observed to be extremely high, yielding a decontamination factor (DF) for a 20-cm-deep bed of the aged charcoal >1 X 10 15 . The results show that essentially no CH 3 I would escape from a 20-cm-deep bed of the aged charcoal and that the requirement for a DF of 1000 for organic iodides in the EFADS filters would be exceeded by a tremendous margin. With such high DFs, the release of iodine from a 20-cm-deep bed would be virtually impossible to detect. The adsorption capacities observed for the aged charcoal samples approach the theoretical chemisorption capacity of 5 wt% TEDA charcoal, indicating that aging in the EFADS for 4 yr has had a negligible impact on the adsorption capacity. The results indicate that the short- and long-term performances of the aged charcoal in the EFADS of Bruce-A following an accident would still far exceed performance requirements. (author)

Projecting Dynastic Majesty: State Ceremony in the Reign of Robert Bruce

Directory of Open Access Journals (Sweden)

Lucinda Dean

2015-07-01

Full Text Available Following the murder of his rival John Comyn on 10 February at Greyfriars in Dumfries, and the crisis this act incited, Robert the Bruce’s inaugural ceremony took place at Scone in late March 1306. Much about this ceremony is speculative; however, subsequent retrospective legitimisation of the Bruce claims to the royal succession would suggest that all possible means by which Robert’s inauguration could emulate his Canmore predecessors and outline his right to rule on a level playing field with his contemporaries were amplified, particularly where they served the common purpose of legitimising Robert’s highly questioned hold on power. Fourteenth-century Scottish history is inextricably entwined in the Wars of Independence, civil strife and an accelerated struggle for autonomous rule and independence. The historiography of this period is unsurprisingly heavily dominated by such themes and, while this has been offset by works exploring subjects such as the tomb of Bruce and the piety of the Bruce dynasty, the ceremonial history of this era remains firmly in the shadows. This paper will address three key ceremonies through which a king would, traditionally, make powerful statements of royal authority: the inauguration or coronation of Bruce; the marriage of his infant son to the English princess Joan of the Tower in 1328, and his extravagant funeral ceremony in 1329. By focusing thus this paper hopes to shed new light on the ‘dark and drublie days’ of fourteenth-century Scotland and reveal that glory, dynastic majesty and pleasure were as central to the Scottish monarchy in this era as war and political turbulence.

Steam generator cleaning campaigns at Bruce A: 1993-1996

International Nuclear Information System (INIS)

Puzzuoli, F.V.; Leinonen, P.J.; Lowe, G.A.

1997-01-01

Boiler chemical cleaning (BOCC) and high-pressure water lancing operations were performed during the Bruce A 1993 Unit 3, 1994 Unit 3, 1995 Unit 1 and 1996 Unit 3 outages to remove secondary side deposits. High-pressure water lancing focused on three boiler areas: tube support plates, to remove broached hole deposits, hot leg U-bend supports to dislodge deposits contributing to boiler tube stress corrosion cracking and tube sheets with the aim of removing accumulated sludge piles and post BOCC insoluble residues. The chemical cleaning processes applied were modified versions of the one developed by the Electric Power Research Institute/Steam Generator Owners Group. During these BOCC operations, corrosion for several key boiler materials was monitored and was well below the specified allowances

Reactor noise analysis applications in NPP I and C systems

Energy Technology Data Exchange (ETDEWEB)

Gloeckler, O. [International Atomic Energy Agency, Wagramer Strosse 5, A-1400 Vienna, Austria Ontario Power Generation, 230 Westney Road South, Ajax, Ont. L1S 7R3 (Canada)

2006-07-01

Reactor noise analysis techniques are used in many NPPs on a routine basis as 'inspection tools' to get information on the dynamics of reactor processes and their instrumentation in a passive, non-intrusive way. The paper discusses some of the tasks and requirements an NPP has to take to implement and to use the full advantages of reactor noise analysis techniques. Typical signal noise analysis applications developed for the monitoring of the reactor shutdown system and control system instrumentation of the Candu units of Ontario Power Generation and Bruce Power are also presented. (authors)

Steam generator replacement at Bruce A: approach, results, and lessons learned

International Nuclear Information System (INIS)

Tomkiewicz, W.; Savage, B.; Smith, J.

2008-01-01

Steam Generator Replacement is now complete in Bruce A Units 1 and 2. In each reactor, eight steam generators were replaced; these were the first CANDU steam generator replacements performed anywhere in the world. The plans for replacement were developed in 2004 and 2005, and were summarized in an earlier paper for the CNS Conference held in November, 2006. The present paper briefly summarizes the methodologies and special processes used such as metrology, cutting and welding and heavy lifting. The paper provides an update since the earlier report and focuses on the project achievements to date, such as: - A combination of engineered methodology, laser metrology and precise remote machining led to accurate first time fit-ups of each new replacement steam generator and steam drums - Lessons learned in the first unit led to schedule improvements in the second unit - Dose received was lowest recorded for any steam generator replacement project. The experience gained and lessons learned from Units 1 and 2 will be valuable in planning and executing future replacement steam generator projects. A video was presented

Bruce and Darlington power pulse and pressure tube integrity programs -status 1995

Energy Technology Data Exchange (ETDEWEB)

Field, G J [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Wylie, J [Ontario Hydro, Tiverton, ON (Canada). Bruce Nuclear Generating Station-A

1996-12-31

The optimum solution to pressure tube fretting at the inlet of the Bruce and Darlington channels, a concern which became very serious following inspections in early 1992, is to remove the inlet bundle and operate with a 12 fuel bundle channel. During analysis of this operating mode a `power pulse` was identified which could occur during an inlet header break where all the fuel in the channel moved rapidly to the inlet of the channel. The pulse was unacceptable and the units were derated until solutions could be implemented. A number of solutions were identified and each station has begun implementation of their specific solution. Implementation has not been without problems and this paper provides a status report on the progress to date of the long bundle implementation solution for Bruce B and Darlington and the fuelling with the flow solution being implemented at Bruce A. Both types of solution have a significant impact on the original concern, fretting of the pressure tube. (author). 1 ref., 6 figs.

AECB staff annual report of Bruce A NGS for the year 1991

International Nuclear Information System (INIS)

1992-11-01

In this report on Bruce A operations during 1991, AECB staff itemizes non-compliances with the operating licence. Non of the violations that occurred at Bruce A resulted in any significant threat to public safety or well-being. There were no exposures of workers to radiation in excess of the regulatory requirements; however, there have been instances of uncontrolled contaminated areas and spread of contamination in the station. Releases of radioactive material to the environment were much below target. The performance of the four special safety systems has been good, with the exception of shutdown system number two on Unit 3. A review of significant event reports and their causes has revealed an apparent lack of a system by which operations and maintenance work is verified as having been carried out correctly. There is a large backlog of maintenance work. Initiatives have been taken to correct this problem. Two important safety issues are discussed in detail. These are the chronic problem of leaking boiler tubes, and the potentially serious problem of fret marks on pressure tubes caused by abnormal fuel support. (Author)

AECB staff annual report of Bruce Heavy Water Plant operation for the year 1991

International Nuclear Information System (INIS)

1992-11-01

Bruce Heavy Water Plant operation was acceptably safe in 1991. There were no breaches of any of the regulations issued under the authority of the Atomic Energy Control Act. There was one violation of the operating licence. For one hour on October 30, 1991, water leaving the plant contained more hydrogen sulphide than Ontario regulations allow. There was no threat to public health or safety or harm to the environment as a result of this violation. One worker was overcome by hydrogen sulphide. The worker did not lose consciousness, but had the symptoms of H 2 S poisoning. Ontario Hydro took actions to increase awareness of the Operating Policy and Principles at Bruce Heavy Water Plant during 1991. All personnel attended a training course, and Ontario Hydro is reviewing all Bruce Heavy Water Plant documentation to ensure it is consistent with the Operating Policies and Principles. Ontario Hydro met 13 of 15 safety-related system availability targets. The AECB is satisfied appropriate action is being taken to improve the performance of the other two systems. Ontario Hydro continued to put heavy emphasis on safety training; however, they did not meet some of their other training targets. Ontario Hydro completed all of the planned emergency exercises at Bruce Heavy Water Plant in 1991. (Author)

R. Bruce Merrifield and Solid-Phase Peptide Synthesis: A Historical Assessment

Energy Technology Data Exchange (ETDEWEB)

Mitchell, A R

2007-12-04

Bruce Merrifield, trained as a biochemist, had to address three major challenges related to the development and acceptance of solid-phase peptide synthesis (SPPS). The challenges were (1) to reduce the concept of peptide synthesis on a insoluble support to practice, (2) overcome the resistance of synthetic chemists to this novel approach, and (3) establish that a biochemist had the scientific credentials to effect the proposed revolutionary change in chemical synthesis. How these challenges were met is discussed in this article.

After reliability centred maintenance. Preventive maintenance living program implementation at Bruce Power

International Nuclear Information System (INIS)

Harazim, Michael L.; Ferguson, Brian J.

2003-01-01

Industrial preventive maintenance (PM) programs represent a large part of plant O and M costs. PM Optimization (PMO) projects represent an effective mechanism for identifying unnecessary PM, extending PM intervals and infusing predictive maintenance (PdM) methods. However, once optimized, what process prevents the PM program from returning to a state of disarray? This is the function of a PM living program (PMLP). In 1997, an independent performance assessment identified concerns with the applicability and effectiveness of all Ontario Power Generation, Inc. (OPGI) PM programs. In response, OPGI instituted an Integrated Maintenance Program (IMP) including Reliability Centred Maintenance (RCM) and a PMLP. It should be noted that the PMLP was developed for the 3 OPGI nuclear Sites (i.e. Bruce, Pickering, and Darlington). Effective 1 May 2001, the Bruce Site has been leased to a group of investors lead by British Energy. This paper is written in historical context and therefore refers to the Bruce Site as part of OPGI. The PMLP is made up of five elements: 1) process control, 2) change control, 3) worker feedback, 4) program performance metrics, and 5) deferral module. A PMLP software tool, originally applied to Duke Energy nuclear plants, was enhanced and customized specifically for the OPGI PMLP, and then implemented at all three of OPGI's nuclear sites. The objective of the OPGI PMLP was to: Provide processes/procedures for continual optimization of all site PM tasks, Ensure effective and timely revision of PM tasks in the work management system, Ensure PM tasks remain applicable/effective at all times, Maintain and enhance PM consistency on a component, system and Site basis, Ensure that new predictive maintenance techniques are applied and integrated with the PM program, Ensure that mandated PM tasks are identified and executed, Provide a mechanism for craft feedback, Meet regulatory requirements for PM program effectiveness, and Provide PM task deferral

AECB staff review of Bruce NGS 'A' operation for the year 1989

International Nuclear Information System (INIS)

1990-06-01

The operation of the Bruce Nuclear Generating Station 'B' is monitored and licensing requirements are enforced by the Atomic Energy Control Board (AECB). This report records the conclusions of the AECB staff assessment of Bruce NGS 'A' during 1989 and the early part of 1990. Overall operation of the station met acceptable safety standards. Despite numerous problems and technical difficulties encountered, station management and supervisory personnel acted with due caution and made decisions in the interests of safety. There was evidence of improvement in a number of key areas, supported by pertinent indicators in the objective measures table. The extensive inspection and maintenance programs carried out during the year revealed the extent of component deterioration due to aging to be larger than expected. Hydrogen embrittlement of pressure tubes, erosion/corrosion of steam and feed water valves, heat exchanger tubes and piping, fouling of boilers and heat exchangers, and environmental damage of electrical equipment are examples. Continued aging of plant equipment and its potential for reducing the margins for safe operation must be taken into account by Ontario Hydro in establishing priorities and target dates for completion of actions to resolve identified problems at Bruce NGS 'A'. (2 tabs.)

Primary separator replacement for Bruce Unit 8 steam generators

International Nuclear Information System (INIS)

Roy, S.B.; Mewdell, C.G.; Schneider, W.G.

2000-01-01

During a scheduled maintenance outage of Bruce Unit 8 in 1998, it was discovered that the majority of the original primary steam separators were damaged in two steam generators. The Bruce B steam generators are equipped with GXP type primary cyclone separators of B and W supply. There were localized perforations in the upper part of the separators and large areas of generalized wall thinning. The degradation was indicative of a flow related erosion corrosion mechanism. Although the unit- restart was justified, it was obvious that corrective actions would be necessary because of the number of separators affected and the extent of the degradation. Repair was not considered to be a practical option and it was decided to replace the separators, as required, in Unit 8 steam generators during an advanced scheduled outage. GXP separators were selected for replacement to minimize the impact on steam generator operating characteristics and analysis. The material of construction was upgraded from the original carbon steel to stainless steel to maximize the assurance of full life. The replacement of the separators was a first of a kind operation not only for Ontario Power Generation and B and W but also for all CANDU plants. The paper describes the degradations observed and the assessments, the operating experience, manufacture and installation of the replacement separators. During routine inspection in 1998, many of the primary steam separators in two of steam generators at Bruce Nuclear Division B Unit 8 were observed to have through wall perforations. This paper describes assessment of this condition. It also discusses the manufacture and testing of replacement primary steam separators and the development and execution of the replacement separator installation program. (author)

Fuel channel design improvements for large CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Villamagna, A; Price, E G; Field, G J [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1996-12-31

From the initial designs used in NPD and Douglas point reactors, the CANDU fuel channel and its components have undergone considerable development. Two major designs have evolved: the Pickering/CANDU 6 design which has 12 fuel bundles in the core and where the new fuel is inserted into the inlet end, and the Bruce/Darlington design which has 13 bundles in the channel and where new fuel is inserted into the outlet end. In the development of a single unit CANDU reactor of the size of a Bruce or Darlington unit which would use a Darlington design calandria, the decision has been made to use the CANDU 6 fuel channel rather than the Darlington design. The CANDU 6 channel has provided excellent performance and will not encounter the degree of maintenance required for the Bruce/Darlington design. The channel design in turn influences the fuelling machine/fuel handling concepts required. The changes to the CANDU 6 fuel channel design to incorporate it in the large unit are small. In fact, the changes that are proposed relate to the desire to increase margins between pressure tube properties and design conditions or ameliorate the consequences of postulated accident conditions, rather than necessary adaptation to the larger unit. Better properties have been achieved in the pressure tube material resulting from alloy development program over the past 10 years. Pressure tubes can now he made with very low hydrogen concentrations so that the hydrogen picked up as deuterium will not exceed the terminal solid solubility for the in-core region in 30 years. The improvements in metal chemistry allow the production of high toughness tubes that retain a high level of toughness during service. A small increase in wall thickness will reduce the dimensional changes without significantly affecting burnup. Changes to increase safety margins from postulated accidents are concentrated on containing the consequences of pressure tube damage. The changes are concentrated on the calandria tube

Life extension of CANDU reactor cores

International Nuclear Information System (INIS)

Millard, J.; Kerker, J.; Albert, M.

2011-01-01

Candu Energy (formerly AECL), in partnership with station operators, has developed a robust methodology for demonstrating the fitness of reactor core structures, and associated reactivity control devices, as an essential element in conducting a station life extension project. The ageing of reactors is affected by ageing mechanisms impacted by operational history and design related factors such as materials, chemistries and stress distributions. The methodology of this life extension work is based on the IAEA TECDOC 1197; which documents practices for ageing management in CANDU reactors. This paper uses the work in Bruce Units 1 and 2, conducted from 2007 through to 2011, to explain the methodology. The work started with analysis of historical operational conditions and identification of the forms of degradation that could have occurred. The assessment and related inspections considered the safety and pressure boundary significance of each item, as well as its failure modes and margins. It then moved through both general and local inspection, focused mainly inside the calandria vessel once the calandria tubes were removed. The inspection found the bulk of the hardware to be in good condition, with a small number of remediation opportunities. In the course of that remediation some foreign material was sampled and removed. The minor remediation was successful and the work was completed through formal documentation of the fitness for extended life. It has been demonstrated through these analyses and visual inspections that the reactor structures and components inspected are free of indications and active degradation mechanisms that would prevent the safe and reliable operation of Bruce A Units 1 and 2 through its next 25 years of life. (author)

Mitigation of organically bound sulphate from water treatment plants at Bruce NGS and impact on steam generator secondary side chemistry control

Energy Technology Data Exchange (ETDEWEB)

Nashiem, R.; Davloor, R.; Harper, B.; Smith, K. [Bruce Power, Tiverton, Ontario (Canada); Gauthier, C. [CTGIX Services Inc., Burlington, Ontario (Canada); Schexnailder, S. [GE Water and Process Technologies, Dallas, Texas (United States)

2010-07-01

Bruce Power is the source of more than 20 per cent of Ontario's electricity and currently operates six reactor units at the Bruce Nuclear Generating Station A (two units) and B (four units) stations located on Lake Huron. This paper discusses the challenges faced and operating experience (OPEX) gained in meeting WANO 1.0 chemistry performance objectives for steam generator secondary side chemistry control, particularly with control of steam generator sulphates. A detailed sampling and analysis program conducted as part of this study concluded that a major contributor to steam generator (SG) elevated sulphates is Organically Bound Sulphate (OBS) in Water Treatment Plants (WTP) effluent. The Bruce A and B WTPs consist of clarification with downstream sand and carbon filtration for Lake Water pre-treatment, which are followed by conventional Ion Exchange (IX) demineralization. Samples taken from various locations in the process stream were analyzed for a variety of parameters including both organic bound and inorganic forms of sulphate. The results are inconclusive with respect to finding the definitive source of OBS. This is primarily due to the condition that the OBS in the samples, which are in relatively low levels, are masked during chemical analysis by the considerably higher inorganic sulphate background. Additionally, it was also determined that on-line Total Organic Carbon (TOC) levels at different WTP locations did not always correlate well with OBS levels in the effluent, such that TOC could not be effectively used as a control parameter to improve OBS performance of the WTP operation. Improvement efforts at both plants focused on a number of areas including optimization of clarifier operation, replacement of IX resins, addition of downstream mobile polishing trailers, testing of new resins and adsorbents, pilot-scale testing with a Reverse Osmosis (RO) rig, review of resin regeneration and backwashing practices, and operating procedure improvements

Mitigation of organically bound sulphate from water treatment plants at Bruce NGS and impact on steam generator secondary side chemistry control

International Nuclear Information System (INIS)

Nashiem, R.; Davloor, R.; Harper, B.; Smith, K.; Gauthier, C.; Schexnailder, S.

2010-01-01

Bruce Power is the source of more than 20 per cent of Ontario's electricity and currently operates six reactor units at the Bruce Nuclear Generating Station A (two units) and B (four units) stations located on Lake Huron. This paper discusses the challenges faced and operating experience (OPEX) gained in meeting WANO 1.0 chemistry performance objectives for steam generator secondary side chemistry control, particularly with control of steam generator sulphates. A detailed sampling and analysis program conducted as part of this study concluded that a major contributor to steam generator (SG) elevated sulphates is Organically Bound Sulphate (OBS) in Water Treatment Plants (WTP) effluent. The Bruce A and B WTPs consist of clarification with downstream sand and carbon filtration for Lake Water pre-treatment, which are followed by conventional Ion Exchange (IX) demineralization. Samples taken from various locations in the process stream were analyzed for a variety of parameters including both organic bound and inorganic forms of sulphate. The results are inconclusive with respect to finding the definitive source of OBS. This is primarily due to the condition that the OBS in the samples, which are in relatively low levels, are masked during chemical analysis by the considerably higher inorganic sulphate background. Additionally, it was also determined that on-line Total Organic Carbon (TOC) levels at different WTP locations did not always correlate well with OBS levels in the effluent, such that TOC could not be effectively used as a control parameter to improve OBS performance of the WTP operation. Improvement efforts at both plants focused on a number of areas including optimization of clarifier operation, replacement of IX resins, addition of downstream mobile polishing trailers, testing of new resins and adsorbents, pilot-scale testing with a Reverse Osmosis (RO) rig, review of resin regeneration and backwashing practices, and operating procedure improvements

Improved synthesis of (3E,6Z,9Z)-1,3,6,9-nonadecatetraene, attraction inhibitor of bruce spanworm, Operophtera bruceata, to pheromone traps for monitoring winter moth, Operophtera brumata.

Science.gov (United States)

Khrimian, Ashot; Lance, David R; Mastro, Victor C; Elkinton, Joseph S

2010-02-10

The winter moth, Operophtera brumata (Lepidoptera: Geometridae), is an early-season defoliator that attacks a wide variety of hardwoods and, in some cases, conifers. The insect is native to Europe but has become established in at least three areas of North America including southeastern New England. The female-produced sex attractant pheromone of the winter moth was identified as (3Z,6Z,9Z)-1,3,6,9-nonadecatetraene (1), which also attracts a native congener, the Bruce spanworm, Operophtera bruceata . Dissection, or (for certainty) DNA molecular testing, is required to differentiate between males of the two species. Thus, a trapping method that is selective for winter moth would be desirable. A geometric isomer of the pheromone, (3E,6Z,9Z)-1,3,6,9-nonadecatetraene (2), can reportedly inhibit attraction of Bruce spanworm to traps without affecting winter moth catch, but use of the pheromone and inhibitor together has not been optimized, nor has the synthesis of the inhibitor. This paper presents two new syntheses of the inhibitor (3E,6Z,9Z)-1,3,6,9-nonadecatetraene based on the intermediate (3Z,6Z)-3,6-hexadecadien-1-ol (4), which has also been utilized in the synthesis of the pheromone. The syntheses combine traditional acetylenic chemistry and Wittig olefination reactions. In one approach, 2 was synthesized in 80% purity (20% being pheromone 1), and in the second, tetraene 2 of 96% purity (and free of 1) was produced in 25% overall yield from dienol 4. The last method benefitted from a refined TEMPO-mediated PhI(OAc)(2) oxidation of 4 and a two-carbon homologation of the corresponding aldehyde 7.

Environmental health scoping study at Bruce Heavy Water Plant

International Nuclear Information System (INIS)

Prior, M.; Mostrom, M.; Coppock, R.; Florence, Z.

1995-10-01

There are concerns that hydrogen sulfide released from the Heavy Water Plant near Kincardine, Ontario may be the cause of the mortalities and morbidities observed in a nearby flock of sheep. The Philosopher's Wool sheep farm is about four kilometres south-southeast of the Bruce Heavy Water Plant. Ontario Hydro, the owner and operator of the Bruce Heavy Water Plant, claims that hydrogen sulphide emissions from the Bruce Heavy Water Plant are within regulatory limits and well below levels that cause harm. Accordingly, the Atomic Energy Control Board commissioned the Alberta Environmental Centre, Alberta Department of Environmental Protection, to develop a scoping study for this environmental health issue. The first objective was to describe a field investigation model to define clearly the environmental health and operation of the sheep farm. The second objective was to describe possible exposure patterns and develop a holistic environmental pathway model. If appropriate, the third study objective was to describe animal models of the actual situation to elucidate specific aspects of the environmental health concerns. It was not the objective of this report to provide a definitive answer to the present environmental health issue. Ontario Hydro provided data to the Alberta Environmental Centre, as di the sheep farmer, the attending veterinarian, the University of Guelph study team, and the Atomic Energy Control Board. A six-tiered strategy of sequential evaluations of the ovine health problem is based on the multiple-response paradigm. It assumes the observed ovine health results are the result of multiple effector events. Each tier constitutes a separate, but inter-related, study. Sequential evaluation and feedback of each tier allow sound scientific judgements and efficient use of resources. (author). 59 refs., 11 tabs., 22 figs

Leak detection capability in CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Azer, N.; Barber, D.H.; Boucher, P.J. [and others

1997-04-01

This paper addresses the moisture leak detection capability of Ontario Hydro CANDU reactors which has been demonstrated by performing tests on the reactor. The tests confirmed the response of the annulus gas system (AGS) to the presence of moisture injected to simulate a pressure tube leak and also confirmed the dew point response assumed in leak before break assessments. The tests were performed on Bruce A Unit 4 by injecting known and controlled rates of heavy water vapor. To avoid condensation during test conditions, the amount of moisture which could be injected was small (2-3.5 g/hr). The test response demonstrated that the AGS is capable of detecting and annunciating small leaks. Thus confidence is provided that it would alarm for a growing pressure tube leak where the leak rate is expected to increase to kg/hr rapidly. The measured dew point response was close to that predicted by analysis.

Leak detection capability in CANDU reactors

International Nuclear Information System (INIS)

Azer, N.; Barber, D.H.; Boucher, P.J.

1997-01-01

This paper addresses the moisture leak detection capability of Ontario Hydro CANDU reactors which has been demonstrated by performing tests on the reactor. The tests confirmed the response of the annulus gas system (AGS) to the presence of moisture injected to simulate a pressure tube leak and also confirmed the dew point response assumed in leak before break assessments. The tests were performed on Bruce A Unit 4 by injecting known and controlled rates of heavy water vapor. To avoid condensation during test conditions, the amount of moisture which could be injected was small (2-3.5 g/hr). The test response demonstrated that the AGS is capable of detecting and annunciating small leaks. Thus confidence is provided that it would alarm for a growing pressure tube leak where the leak rate is expected to increase to kg/hr rapidly. The measured dew point response was close to that predicted by analysis

Validation of DRAGON side-step method for Bruce-A restart Phase-B physics tests

International Nuclear Information System (INIS)

Shen, W.; Ngo-Trong, C.; Davis, R.S.

2004-01-01

The DRAGON side-step method, developed at AECL, has a number of advantages over the all-DRAGON method that was used before. It is now the qualified method for reactivity-device calculations. Although the side-step-method-generated incremental cross sections have been validated against those previously calculated with the all-DRAGON method, it is highly desirable to validate the side-step method against device-worth measurements in power reactors directly. In this paper, the DRAGON side-step method was validated by comparison with the device-calibration measurements made in Bruce-A NGS Unit 4 restart Phase-B commissioning in 2003. The validation exercise showed excellent results, with the DRAGON code overestimating the measured ZCR worth by ∼5%. A sensitivity study was also performed in this paper to assess the effect of various DRAGON modelling techniques on the incremental cross sections. The assessment shows that the refinement of meshes in 3-D and the use of the side-step method are two major reasons contributing to the improved agreement between the calculated ZCR worths and the measurements. Use of different DRAGON versions, DRAGON libraries, local-parameter core conditions, and weighting techniques for the homogenization of tube clusters inside the ZCR have a very small effect on the ZCR incremental thermal absorption cross section and ZCR reactivity worth. (author)

Integrated inspection programs at Bruce Heavy Water Plant

Energy Technology Data Exchange (ETDEWEB)

Brown, K C [Ontario Hydro, Tiverton, ON (Canada)

1993-12-31

Quality pressure boundary maintenance and an excellent loss prevention record at Bruce Heavy Water Plant are the results of the Material and Inspection Unit`s five inspection programs. Experienced inspectors are responsible for the integrity of the pressure boundary in their own operating area. Inspectors are part of the Technical Section, and along with unit engineering staff, they provide technical input before, during, and after the job. How these programs are completed, and the results achieved, are discussed. 5 figs., 1 appendix.

Integrated inspection programs at Bruce Heavy Water Plant

International Nuclear Information System (INIS)

Brown, K.C.

1992-01-01

Quality pressure boundary maintenance and an excellent loss prevention record at Bruce Heavy Water Plant are the results of the Material and Inspection Unit's five inspection programs. Experienced inspectors are responsible for the integrity of the pressure boundary in their own operating area. Inspectors are part of the Technical Section, and along with unit engineering staff, they provide technical input before, during, and after the job. How these programs are completed, and the results achieved, are discussed. 5 figs., 1 appendix

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)

Design of the MiniSLAR system for Bruce A

International Nuclear Information System (INIS)

Gray, M.G.

1995-01-01

Cancellation of Bruce A Retube created the need to perform SLAR on Unit 1. The existing SLAR system cannot reach Unit 1 and alternative systems had limitations. The concept and design of MiniSLAR were driven by the availability of existing components made for Retube. The MiniSLAR concept was developed by a team with members representing operators, technicians, and designers from various departments within Ontario Hydro and GE Canada. Overall project leadership was provided by Bruce A Projects and Modifications Department with assistance from Ontario Hydro Nuclear Technology Services. The responsibility for detailed design was assigned by Ontario Hydro to GE Canada. The detailed design proceeded with continual input and review by the team. The MiniSLAR delivery machine consists of a closure removal ram, a shield plug removal ram and a SLAR tool delivery ram attached to the sliding plate of a horizontal indexing mechanism. The moving plate is constrained by guide rails to a fixed plate and seals against it with o-rings. A snout and clamping mechanism mounts on the front of the fixed plate. The machine mounts atop a work table which provides the various motions required for endfitting engagement. Some operations are performed manually while others are remote and automatic. (author)

An appreciation of Bruce and Young's (1986) serial stage model of face naming after 25 years.

Science.gov (United States)

Hanley, J Richard

2011-11-01

The current status of Bruce and Young's (1986) serial model of face naming is discussed 25 years after its original publication. In the first part of the paper, evidence for and against the serial model is reviewed. It is argued that there is no compelling reason why we should abandon Bruce and Young's claim that recall of a name is contingent upon prior retrieval of semantic information about the person. The current status of the claim that people's names are more difficult to recall than the names of objects is then evaluated. Finally, an account of the anatomical location in the brain of Bruce and Young's three processing stages (face familiarity, retrieval of semantic information, retrieval of names) is suggested. In particular, there is evidence that biographical knowledge about familiar people is stored in the right anterior temporal lobes (ATL) and that the left temporal pole (TP) is heavily involved in retrieval of the names of familiar people. The issue of whether these brain areas play a similar role in object processing is also discussed. ©2011 The British Psychological Society.

The status of improved pressurized heavy water reactor development - A new option for PHWR -

International Nuclear Information System (INIS)

Park, Tae Keun; Yeo, Ji Won

1996-03-01

Currently, the 900 MWe class Improved Pressurized Heavy Water Reactor (PHWR), which is a type of CANDU reactor based on the systems and components of operating CANDU plants, is under development. The improved PHWR has a 480 fuel channel calandria, uses 37 element natural uranium fuel bundles and has a single unit containment. Adaptation of a steel-lined containment structure and improved containment isolation systems permit a reduced exclusion area boundary (EAB) compared to the existing larger capacity CANDU reactors (Darlington, Bruce B). The improved PHWR buildings are arranged to achieve minimum spacing between reactor units. Plant safety and economy are increased through various design changes based on the operating experience of existing CANDU plants. 4 refs. (Author)

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)

Analysis of log rate noise in Ontario's CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Hinds, H.W. [Dynamic Simulation and Analysis Corp., Deep River, Ontario (Canada); Banica, C.; Arguner, D. [Ontario Power Generation, Ajax, Ontario (Canada); Scharfenberg, R. [Bruce Power, Tiverton, Ontario (Canada)

2007-07-01

In the fall of 2003, the operators noticed that in the recently-refurbished Bruce A Shutdown System no. 1 (SDS1) the noise level in Log Rate signals were much larger than before. At the request of the Canadian Nuclear Safety Commission (CNSC), all Canadian CANDU reactors took action to characterize their Log Rate noise. Staff of the Inspection and Maintenance Services division of Ontario Power Generation (OPG) has collected high-speed high-accuracy noise data from nearly all 16 Ontario reactors, either as part of routine measurements before planned outages or as a dedicated noise recording. This paper gives the results of examining a suitable subset of this data, with respect to the characteristics and possible causes of Log Rate noise. The reactor and instrumentation design is different at each station: the locations of the moderator injection nozzles, the location of the ion chambers for each system, and the design of the Log Rate amplifiers. It was found that the Log noise (source of Log Rate noise) was much larger for those ion chambers in the path of the moderator injection nozzles, compared to those which were not in the path. This 'extra' Log noise would then be either attenuated or amplified depending on the transfer function (time constants) of the Log Rate amplifier. It was also observed that most of the Log and Log Rate noise is independent of any other signal measured. Although all CANDU reactors in Ontario have Log and Log Rate noise, the Bruce A SDS1 system has the largest amount of Log Rate noise, because (a) its SDS1 (and RRS) ion chambers are at the top of the reactor in the path of the moderator injection nozzles, and (b) its SDS1 Log Rate amplifiers have the smallest time constants. (author)

Reactor core flow measurements during plant start-up using non-intrusive flow meter CROSSFLOW

Energy Technology Data Exchange (ETDEWEB)

Kanda, V.; Sharp, B.; Gurevich, A., E-mail: vkanda@amag-inc.com, E-mail: bsharp@amag-inc.com, E-mail: agurevich@amag-inc.com [Advanced Measurement & Analysis Group Inc., Ontario (Canada); Gurevich, Y., E-mail: yuri.gurevich@daystartech.ca [Daystar Technologies Inc., Ontario (Canada); Selvaratnarajah, S.; Lopez, A., E-mail: sselvaratnarajah@amag-inc.com, E-mail: alopez@amag-inc.com [Advanced Measurement & Analysis Group Inc., Ontario (Canada)

2013-07-01

For the first time, direct measurements of the total reactor coolant flow and the flow distribution between the inner reactor zone and the outer zone were conducted using the non-intrusive clamp on ultrasonic cross-correlation flow meter, CROSSFLOW, developed and manufactured by Advanced Measurement & Analysis Group Inc. (AMAG). The measurements were performed at Bruce Power A Unit 1 on the Pump Discharge piping of the Primary Heat Transport (PHT) system during start-up. This paper describes installation processes, hydraulic testing, uncertainty analysis and traceability of the measurements to certified standards. (author)

Estimating the number of latent cracks in pressure tube joints at Bruce unit 2

International Nuclear Information System (INIS)

Schwarz, C.J.

1983-10-01

A model was built to estimate the number of hydride cracks which might have arisen in the rolled joints of Bruce unit 2 prior to the stress relieving operation. The model estimated that about 100 such cracks might exist. Since this estimate is based on experiments that were thermally cycled and since cycling did not occur in Bruce, prior to stress relieving the actual number is expected to be substantially lower. A sensitivity analysis of the model showed that it is sensitive to the assumptions of stress levels, probability of initiation and distribution of initiation time. A better estimate could be made if more data were available on these parameters under realistic conditions. Therefore, the recommendation is made to collect more information about these factors under realistic conditions

Exercise capacity in Dutch children : New reference values for the Bruce treadmill protocol

NARCIS (Netherlands)

M.H.M. van der Cammen-van Zijp (Monique); H.J.G. van den Berg-Emons (Rita); S.P. Willemsen (Sten); H.J. Stam (Henk); D. Tibboel (Dick); H. IJsselstijn (Hanneke)

2010-01-01

textabstractThe Bruce treadmill protocol is suitable for children 4 years of age and older. Dutch reference values were established in 1987. We considered that children's exercise capacity has deteriorated due to changes in physical activity patterns and eating habits. We determined new reference

SORO post-simulations of Bruce A Unit 4 in-core flux detector verification tests

Energy Technology Data Exchange (ETDEWEB)

Braverman, E.; Nainer, O. [Bruce Power, Nuclear Safety Analysis and Support Dept., Toronto, Ontario (Canada)]. E-mail: Evgeny.Braverman@brucepower.com; Ovidiu.Nainer@brucepower.com

2004-07-01

During the plant equipment assessment prior to requesting approval for restart of Bruce A Units 3 and 4 it was determined that all in-core flux detectors needed to be replaced. Flux detector verification tests were performed to confirm that the newly installed detectors had been positioned according to design specifications and that their response closely follows the calculated flux shape changes caused by selected reactivity mechanism movements. By comparing the measured and post-simulated RRS and NOP detector responses to various perturbations, it was confirmed that the new detectors are wired and positioned correctly. (author)

Bruce Unit 2 lay-up engineering assessment

International Nuclear Information System (INIS)

Iley, D.

1995-01-01

The overall lay-up program initiated as a result of the strategic decision to shut down Bruce A unit 2 is briefly described as an introduction to the engineering assessment of the unit 2 systems. The assessment has identified the need to prepare 67 system and 9 equipment lay-up specifications. A summary of the selected system specifications is described. A complete summary and the specifications and the status of unit 2 systems and equipment required to support lay-up and/or the other three operating units is available on request due to the volume of the information. Some logistical details of the lay-up implementation plans, results, and problems to date demonstrate the complexity of the lay-up requirements for a nuclear unit in a multi-unit CANDU station. (author)

Restoration to serviceability of Bruce 'A' heat transfer equipment

International Nuclear Information System (INIS)

Gammage, D.; Machowski, C.; McGillivray, R.; Durance, D.; Kazimer, D.; Werner, K.

2009-01-01

Bruce Units 1 to 4 were shut down during the 1990s by the former Ontario Hydro, due in part to a long list of system and equipment deficiencies and concerns, including steam generator tube degradation as a consequence of the then-existing steam generator secondary side water chemistry conditions. Upon its creation in 2001, and following a program of condition assessment, Bruce Power was able to determine that Units 3 and 4 could return to service; but that Units 1 and 2 would require refurbishment. That Refurbishment Program, which is currently well advanced, included the re-assessment of the condition of equipment throughout the plant including the heat transfer equipment; and determination item-by-item as to what inspection, cleaning, repair, or even replacement would be required to put the equipment into a condition where it could be expected to operate reliably for the additional 30 years expected from the plant. Clearly the objective is to suitably restore the equipment to serviceability without doing more refurbishment work than is warranted - without replacing equipment except where absolutely necessary. The first task in such a program is determination of its scope - i.e. a listing of all heat exchangers. That list included everything from the steam generators (which required replacement, now completed), to much smaller heat exchangers in the heavy water upgrader systems (which were found to be in very good overall condition). There is also a very large number of other so-called 'balance-of-plant' heat exchangers; these include the maintenance coolers, moderator heat exchangers, shutdown coolers and a whole raft of smaller coolers - many of which are cooled directly by lake water with its potential for bio-fouling and 'BIC' (Biologically Induced Corrosion). This paper focuses primarily on the engineering assessment, inspection, repair and general refurbishment of the balance-of-plant heat exchangers. As will be discussed in the paper, the assessment of the

Kuldlõvid Louise Bourgeois'le, Bruce Naumanile ja Itaalia paviljonile / Reet Varblane

Index Scriptorium Estoniae

Varblane, Reet, 1952-

1999-01-01

Veneetsia 48. rahvusvahelise kunstibiennaali preemiasaajad, premeeritud tööd, korralduskomitee ja žürii koosseis. Kuldlõvid: Louise Bourgeois, Bruce Nauman, Itaalia paviljon (Monica Bonvicini, Bruna Esposito, Luisa Lambri, Paola Pivi, Grazia Toderi ühisprojekt); kolm rahvusvahelist preemiat: Doug Aitken, Cai Gou-Qiang, Shirin Neshat; žürii tõstis esile: Georges Abeagbo, Eija-Liisa Ahtila, Katarzyna Kozura ? (Kozyra), Lee Bul; UNESCO preemia: Ghada Amer

Chemistry control at Bruce NGS 'B' from constructed to commercial operation

International Nuclear Information System (INIS)

Roberts, J.G.

1987-01-01

Pre-operational storage chemistry and flushing of the secondary side is described. The approach devised for Bruce NGS 'B' Unit 6 was unique for an Ontario Hydro Nuclear Unit. The significance of the improved Construction installation and Quality Assurance procedures, combined with those of Operations is identified. Secondary side chemistry during both commissioning and later operation is reported. It will be shown that the application of ALARA (As Low As is Reasonably Achievable) concept has resulted in tighter chemical specifications being met

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)

Analogue to digital upgrade project-boiler feedwater control system for Bruce Power nuclear units 1 & 2

International Nuclear Information System (INIS)

Long, R.

2012-01-01

Bruce Power Nuclear Generating Station A, “Bruce A” is in the final stages of its Restart Project. This capital project will see a large scale rehabilitation of Units 1 and 2 resulting in addition of 1500MW of safe, reliable, clean electricity to the Ontario grid. Restart Project Scope 375, Boiler Feedwater Controls Upgrade was sanctioned to replace obsolete analog devices with a modern digital control system. This project replaced the existing Foxboro H Line analog controls which comprised of 81 individual control modules and support instrumentation. The replacement system was a Triconex Triple Modular Redundant PLC which interfaces with two redundant touch screen monitors. The upgraded digital system incorporates the following controls: 1. Boiler Level Control Loops 2. Dearator Level Control Loops 3. Dearator Pressure Control Loops 4. Boiler Feedwater Recirculation Flow Control Loops A number of technical challenges were addressed when installing a new digital system within the existing plant configuration. Interfaces to new, old and refurbished field devices must be understood as well as implications of connecting to the plant’s Digital Control Computers (DCC’s) and newly installed Steam Generators. The overall project involved many stakeholders to address various requirements from conceptual / design stage through procurement, construction, commissioning and return to service. In addition, the project highlighted the unique requirements found in Nuclear Industry with respect to Human Factors and Software Quality Assurance. (author)

Steam generator leak detection at Bruce A Unit 1

International Nuclear Information System (INIS)

Maynard, K.J.; McInnes, D.E.; Singh, V.P.

1997-01-01

A new steam generator leak detection system was recently developed and utilized at Bruce A. The equipment is based on standard helium leak detection, with the addition of moisture detection and several other capability improvements. All but 1% of the Unit 1 Boiler 03 tubesheet was inspected, using a sniffer probe which inspected tubes seven at a time and followed by individual tube inspections. The leak search period was completed in approximately 24 hours, following a prerequisite period of several days. No helium leak indications were found anywhere on the boiler. A single water leak indication was found, which was subsequently confirmed as a through-wall defect by eddy current inspection. (author)

Fighting with Reality: Considering Mark Johnson's Pragmatic Realism through Bruce Lee's Jeet Kune Do Method

Science.gov (United States)

Miller, Alexander David

2015-01-01

This dissertation considers the supportive and complementary relation between Mark Johnson's embodied realism and Bruce Lee's Jeet Kune Do as a philosophical practice. In exploring this relationship, the emphasis on one's embodiment condition and its relationship with metaphor and self-expression are the primary focus. First, this work involves…

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)

Impact of the 37M fuel design on reactor physics characteristics

International Nuclear Information System (INIS)

Perez, R.; Ta, P.

2013-01-01

For CANDU nuclear reactors, aging of the Heat Transport System (HTS) leads to, among other effects, a reduction on the Critical Heat Flux (CHF) and dryout margin. In an effort to mitigate the impact of aging of the HTS on safety margins, Bruce Power is introducing a design change to the standard 37-element fuel bundle known as the modified 37-element fuel bundle, or 37M for short. As part of the overall design change process it was necessary to assess the impact of the modified fuel bundle design on key reactor physics parameters. Quantification of this impact on lattice cell properties, core reactivity properties, etc., was reached through a series of calculations using state-of-the-art lattice and core physics models, and comparisons against results for the standard fuel bundle. (author)

Phytotoxicology section investigation in the vicinity of the Bruce Nuclear Power Development, the Pickering Nuclear Generating Station and the Darlington Nuclear Generating Station, in October, 1989

International Nuclear Information System (INIS)

1991-02-01

The Phytotoxicology Section, Air Resources Branch is a participant in the Pickering and Bruce Nuclear Contingency Plans. The Phytotoxicology Emergency Response Team is responsible for collecting vegetation samples in the event of a nuclear emergency at any of the nuclear generating stations in the province. As part of its responsibility the Phytotoxicology Section collects samples around the nuclear generating stations for comparison purposes in the event of an emergency. Because of the limited frequency of sampling, the data from the surveys are not intended to be used as part of a regulatory monitoring program. These data represent an effort by the MOE to begin to establish a data base of tritium concentrations in vegetation. The Phytotoxicology Section has carried out seven surveys in the vicinity of Ontario Hydro nuclear generating stations since 1981. Surveys were conducted for tritium in snow in the vicinity of Bruce Nuclear Power Development (BNPD), February, 1981; tritium in cell-free water of white ash in the vicinity of BNPD, September, 1981; tritium in snow in the vicinity of BNPD, March, 1982; tritium in tree sap in the vicinity of BNPD, April, 1982; tritium in tree sap in the vicinity of BNPD, April, 1984, tritium in the cell-free water of white ash in the vicinity of BNPD, September, 1985; and, tritium in cell-free water of grass in the vicinity of Pickering Nuclear Generation Station (PNGS), October 1986. In all cases a pattern of decreasing tritium levels with increasing distance from the stations was observed. In October, 1989, assessment surveys were conducted around Bruce Nuclear Power Development, the Pickering Nuclear Generating Station and the new Darlington Nuclear Generating Station (DNGS). The purpose of these surveys was to provide baseline data for tritium in cell-free water of grass at all three locations at the same time of year. As none of the reactor units at DNGS had been brought on line at the time of the survey, this data was to be

The design and installation of a core discharge monitor for CANDU-type reactors

International Nuclear Information System (INIS)

Halbig, J.K.; Monticone, A.C.; Ksiezak, L.; Smiltnieks, V.

1990-01-01

A new type of surveillance systems that monitors neutron and gamma radiation in a reactor containment is being installed at the Ontario Hydro Darlington Nuclear Generating Station A, Unit 2. Unlike video or film surveillance that monitors mechanical motion, this system measures fuel-specific radiation emanating from irradiated fuel as it is pushed from the core of CANDU-type reactors. Proof-of-principle measurements have been carried out at Bruce Nuclear Generating Station A, Unit 3. The system uses (γ,n) threshold detectors and ionization detectors. A microprocessor-based electronics package, GRAND-II (Gamma Ray and Neutron Detector electronics package), provides detector bias, preamplifier power, and signal processing. Firmware in the GRAND-2 controls the surveillance activities, including data acquisition and a level of detector authentication, and it handles authenticated communication with a central data logging computer. Data from the GRAND-II are transferred to an MS-DOS-compatible computer and stored. These data are collected and reviewed for fuel-specific radiation signatures from the primary detector and proper ratios of signals from secondary detectors. 5 figs

Review of Ontario Hydro Pickering 'A' and Bruce 'A' nuclear generating stations' accident analyses

International Nuclear Information System (INIS)

Serdula, K.J.

1988-01-01

Deterministic safety analysis for the Pickering 'A' and Bruce 'A' nuclear generating stations were reviewed. The methodology used in the evaluation and assessment was based on the concept of 'N' critical parameters defining an N-dimensional safety parameter space. The reviewed accident analyses were evaluated and assessed based on their demonstrated safety coverage for credible values and trajectories of the critical parameters within this N-dimensional safety parameter space. The reported assessment did not consider probability of occurrence of event. The reviewed analyses were extensive for potential occurrence of accidents under normal steady-state operating conditions. These analyses demonstrated an adequate assurance of safety for the analyzed conditions. However, even for these reactor conditions, items have been identified for consideration of review and/or further study, which would provide a greater assurance of safety in the event of an accident. Accident analyses based on a plant in a normal transient operating state or in an off-normal condition but within the allowable operating envelope are not as extensive. Improvements in demonstrations and/or justifications of safety upon potential occurrence of accidents would provide further assurance of adequacy of safety under these conditions. Some events under these conditions have not been analyzed because of their judged low probability; however, accident analyses in this area should be considered. Recommendations are presented relating to these items; it is also recommended that further study is needed of the Pickering 'A' special safety systems

Uncertainties in gas dispersion at the Bruce heavy water plant

International Nuclear Information System (INIS)

Alp, E.; Ciccone, A.

1995-07-01

There have been concerns regarding the uncertainties in atmospheric dispersion of gases released from the Bruce Heavy Water Plant (BHWP). The concern arises due to the toxic nature of H 2 S, and its combustion product SO 2 . In this study, factors that contribute to the uncertainties, such as the effect of the shoreline setting, the potentially heavy gas nature of H 2 S releases, and concentration fluctuations, have been investigated. The basic physics of each of these issues has been described along with fundamental modelling principles. Recommendations have been provided on available computer models that would be suitable for modelling gas dispersion in the vicinity of the BHWP. (author). 96 refs., 4 tabs., 25 figs

Uncertainties in gas dispersion at the Bruce heavy water plant

Energy Technology Data Exchange (ETDEWEB)

Alp, E; Ciccone, A [Concord Environmental Corp., Downsview, ON (Canada)

1995-07-01

There have been concerns regarding the uncertainties in atmospheric dispersion of gases released from the Bruce Heavy Water Plant (BHWP). The concern arises due to the toxic nature of H{sub 2}S, and its combustion product SO{sub 2}. In this study, factors that contribute to the uncertainties, such as the effect of the shoreline setting, the potentially heavy gas nature of H{sub 2}S releases, and concentration fluctuations, have been investigated. The basic physics of each of these issues has been described along with fundamental modelling principles. Recommendations have been provided on available computer models that would be suitable for modelling gas dispersion in the vicinity of the BHWP. (author). 96 refs., 4 tabs., 25 figs.

Coffee Cups, Canoes, Airplanes and the Lived Experience: Reflections on the Works of Bertram (Chip) Bruce

Science.gov (United States)

Haythornthwaite, Caroline

2014-01-01

A career spent in research, teaching, and engagement with community entails a lifetime of assemblage of meaning from people, resources, technologies and experience. In his work, Bertram (Chip) Bruce has long engaged with how we create such an assemblage of meaning from our formal and found learning, and from the "lived experience" of…

Development of restriction enzyme analyses to distinguish winter moth from bruce spanworm and hybrids between them

Science.gov (United States)

Marinko Sremac; Joseph Elkinton; Adam. Porter

2011-01-01

Elkinton et. al. recently completed a survey of northeastern North America for the newly invasive winter moth, Operophtera brumata L. The survey used traps baited with the winter moth pheromone, which consists of a single compound also used by Bruce spanworm, O. bruceata (Hulst), the North American congener of winter moth. Our...

Bundle 13 position verification tool description and on-reactor use

Energy Technology Data Exchange (ETDEWEB)

Onderwater, T G [Canadian General Electric Co. Ltd., Peterborough, ON (Canada)

1997-12-31

To address the Power Pulse problem, Bruce B uses Gap: a comprehensive monitoring program by the station to maintain the gap between the fuel string and the upstream shield plug. The gap must be maintained within a band. The gap must not be so large as to allow excessive reactivity increases or cause high impact forces during reverse flow events. It should also not be so small as to cause crushed fuel during rapid, differential reactor/fuel string cool downs. Rapid cool downs are infrequent. The Bundle 13 Position Verification Tool (BPV tool) role is to independently measure the position of the upstream bundle of the fuel string. The measurements are made on-reactor, on-power and will allow verification of the Gap Management system`s calculated fuel string position. This paper reviews the reasons for developing the BPV tool. Design issues relevant to safe operation in the fuelling machine, fuel channel and fuel handling equipment are also reviewed. Tests ensuring no adverse effects on channel pressure losses are described and actual on-reactor, on-power results are discussed. (author). 4 figs.

Bundle 13 position verification tool description and on-reactor use

International Nuclear Information System (INIS)

Onderwater, T.G.

1996-01-01

To address the Power Pulse problem, Bruce B uses Gap: a comprehensive monitoring program by the station to maintain the gap between the fuel string and the upstream shield plug. The gap must be maintained within a band. The gap must not be so large as to allow excessive reactivity increases or cause high impact forces during reverse flow events. It should also not be so small as to cause crushed fuel during rapid, differential reactor/fuel string cool downs. Rapid cool downs are infrequent. The Bundle 13 Position Verification Tool (BPV tool) role is to independently measure the position of the upstream bundle of the fuel string. The measurements are made on-reactor, on-power and will allow verification of the Gap Management system's calculated fuel string position. This paper reviews the reasons for developing the BPV tool. Design issues relevant to safe operation in the fuelling machine, fuel channel and fuel handling equipment are also reviewed. Tests ensuring no adverse effects on channel pressure losses are described and actual on-reactor, on-power results are discussed. (author). 4 figs

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).

Adapting Scott and Bruce's General Decision-Making Style Inventory to Patient Decision Making in Provider Choice.

Science.gov (United States)

Fischer, Sophia; Soyez, Katja; Gurtner, Sebastian

2015-05-01

Research testing the concept of decision-making styles in specific contexts such as health care-related choices is missing. Therefore, we examine the contextuality of Scott and Bruce's (1995) General Decision-Making Style Inventory with respect to patient choice situations. Scott and Bruce's scale was adapted for use as a patient decision-making style inventory. In total, 388 German patients who underwent elective joint surgery responded to a questionnaire about their provider choice. Confirmatory factor analyses within 2 independent samples assessed factorial structure, reliability, and validity of the scale. The final 4-dimensional, 13-item patient decision-making style inventory showed satisfactory psychometric properties. Data analyses supported reliability and construct validity. Besides the intuitive, dependent, and avoidant style, a new subdimension, called "comparative" decision-making style, emerged that originated from the rational dimension of the general model. This research provides evidence for the contextuality of decision-making style to specific choice situations. Using a limited set of indicators, this report proposes the patient decision-making style inventory as valid and feasible tool to assess patients' decision propensities. © The Author(s) 2015.

Hydrogeological evidence of low rock mass permeabilities in ordovician strata: Bruce nuclear site

International Nuclear Information System (INIS)

Beauheim, R.L.; Roberts, R.M.; Avis, J.D.; Heagle, D.

2011-01-01

One of the key attributes contributing to the suitability of the Bruce nuclear site to host a Deep Geologic Repository (DGR) for Low and Intermediate Level Waste (L&ILW) is the low permeability of the Ordovician host rock and of the overlying and underlying strata. The permeability of these rocks is so low that diffusion is a much more significant transport mechanism than advection. Hydrogeological evidence for the low permeability of the Ordovician strata comes from two principal sources, direct and indirect. Direct evidence of low permeability is provided by the hydraulic testing performed in deep boreholes, DGR-2 through DGR-6. Straddle-packer hydraulic testing was performed in 57 Ordovician intervals in these five holes. The testing provided continuous coverage using ~30-m straddle intervals of the Ordovician strata exposed in boreholes DGR-2, DGR-3, DGR-4, and DGR-5, while testing was targeted on discontinuous 10.2-m intervals in DGR-6. The average horizontal hydraulic conductivities of these intervals determined from the tests ranged from 2E-16 to 2E-10 m/s. The Lower Member of the Cobourg Formation, which is the proposed host formation for the DGR, was found to have a horizontal hydraulic conductivity of 4E-15 to 3E-14 m/s. The only horizontal hydraulic conductivity values measured that were greater than 2E-12 m/s are from the Black River Group, located at the base of the Ordovician sedimentary sequence. Indirect evidence of low permeability is provided by the observed distribution of hydraulic heads through the Ordovician sequence. Hydraulic head profiles, defined by hydraulic testing and confirmed by Westbay multilevel monitoring systems, show significant underpressures relative to a density-compensated hydrostatic condition throughout most of the Ordovician strata above the Black River Group, whereas the Black River Group is overpressured. Pressure differences of 1 MPa or more are observed between adjacent intervals in the boreholes. The observed

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)

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)

Expected changes in competency assurance at Bruce Power over next decade

Energy Technology Data Exchange (ETDEWEB)

Horton, C., E-mail: chip.horton@brucepower.com [Bruce Power, Tiverton, Ontario (Canada)

2013-07-01

There will be ten expected changes in competency assurance at Bruce Power over next decade. These changes are: outage worker supplemental staff change; entry level power worker union staff; self-paced process & software training; electronic confirmation of qualifications; JIT built into processes for low frequency tasks; more terminal objective back to fundamentals vs fundamentals build to terminal objective instructional design; more learning by practice / doing in a safe environment; SAT emphasis move from analysis, design & development to evaluation & maintenance of task lists (clearly tied to working rights); more formal tacit knowledge capture programs, systematic movement of knowledge from tacit to explicit; more breadth & depth training driven by demographic changes.

Expected changes in competency assurance at Bruce Power over next decade

International Nuclear Information System (INIS)

Horton, C.

2013-01-01

There will be ten expected changes in competency assurance at Bruce Power over next decade. These changes are: outage worker supplemental staff change; entry level power worker union staff; self-paced process & software training; electronic confirmation of qualifications; JIT built into processes for low frequency tasks; more terminal objective back to fundamentals vs fundamentals build to terminal objective instructional design; more learning by practice / doing in a safe environment; SAT emphasis move from analysis, design & development to evaluation & maintenance of task lists (clearly tied to working rights); more formal tacit knowledge capture programs, systematic movement of knowledge from tacit to explicit; more breadth & depth training driven by demographic changes.

Hourly and seasonable variation in catch of winter moths and bruce spanworm in pheromone-baited traps

Science.gov (United States)

Joseph Elkinton; Natalie Leva; George Boettner; Roy Hunkins; Marinko. Sremac

2011-01-01

Elkinton et al. recently completed a survey of northeastern North America for the newly invasive winter moth, Operophtera brumata L. The survey used traps baited with the winter moth pheromone, which, as far as it is known, consists of a single compound that is also used by Bruce spanworm, the North American congener of winter moth, O....

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

Event review: International Knapping Workshop, with Bruce Bradley, Fazenda Monte Alto, Dourado, SP (Brazil

Directory of Open Access Journals (Sweden)

Elisa Theodora Adriana van Veldhuizen

2016-07-01

Full Text Available The event took place from 3 till 8 July 2016 at Fazenda Monte Alto, Dourado, SP, Brazil. The aim of the course was to provide intensive knapping training in order to enhance analytical methods and procedures. This training was not only for students, but also professionals who were interested in the course. The course was given by Bruce Bradley (University of Exeter, who has extensive experience with Stone Age technologies and experimental archaeology. Mercedes Okumura (PPGArq, National Museum, Federal University of Rio de Janeiro and Astolfo G. M. Araujo (Museum of Archaeology and Ethnology, University of São Paulo organized the course, which was sponsored by Fazenda Monte Alto, Café Helena, and the British Academy, Newton Mobility Grants Scheme (NG140077. The workshop had 15 participants from Brazil, Uruguay, the Netherlands and Canada.

Plutonium Consumption Program, CANDU Reactor Project final report

Energy Technology Data Exchange (ETDEWEB)

1994-07-31

DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro`s Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel.

Plutonium Consumption Program, CANDU Reactor Project final report

International Nuclear Information System (INIS)

1994-01-01

DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro's Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel

Two new species of the stenopodidean shrimp genus Spongiocaris Bruce & Baba, 1973 (Crustacea: Decapoda: Spongicolidae) from the Indo-West Pacific.

Science.gov (United States)

Komai, Tomoyuki; Grave, Sammy De; Saito, Tomomi

2016-05-17

Two new species of the deep-water spongicolid genus Spongiocaris Bruce & Baba, 1973, are described and illustrated from two localities in the Indo-West Pacific. Spongiocaris panglao n. sp. is described on the basis of material from the Bohol Sea, the Philippines, at depths of 220-731 m. Spongiocaris tuerkayi n. sp. is described on the basis of material from Atlantis Bank in the southwestern Indian Ocean at depths of 743-1053 m. Among eight known congeners, both new species appear close to S. semiteres Bruce & Baba, 1973, differing in the rostral length and armature, shape of the carapace, telsonal armature, development of the grooming apparatus of the first pereopod and shape of the third pereopod chela. An identification key to the species currently assigned to Spongiocaris is presented.

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)

Evaluation of a pilot fish handling system at Bruce NGS 'A'

International Nuclear Information System (INIS)

Griffiths, J.S.

1985-10-01

A pilot fish recovery system using a Hidrostal fish pump was tested in the Bruce NGS 'A' forebay during June, 1984. Despite low forebay fish concentrations, the system was capable of capturing 97,000 alewife/day (3900 kg) if operated continuously. Post-pumping survival averaged 97%. It is estimated that a single pump could handle alewife runs in the 40,000 to 70,000 kg range, but multiple pumps or a single larger pump would be required to assure station protection from the largest runs (>100,000 kg). Results indicate that tank/trailer return of pumped fish is feasible, but other alternatives for returning fish to Lake Huron are also being considered

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.)

Hydraulic Testing of Silurian and Ordovician Strata at the Bruce Site

Science.gov (United States)

Beauheim, R. L.; Avis, J. D.; Chace, D. A.; Roberts, R. M.; Toll, N. J.

2009-05-01

Ontario Power Generation is proposing a Deep Geologic Repository (DGR) for the long-term management of its Low and Intermediate Level Radioactive Waste (L&ILW) within a Paleozoic-age sedimentary sequence beneath the Bruce Site near Tiverton, Ontario, Canada. The concept envisions that the DGR would be excavated at a depth of approximately 680 m within the Ordovician Cobourg Formation, a massive, dense, argillaceous limestone. A key attribute of the Bruce site is the extremely low permeabilities associated with the thick Ordovician carbonate and argillaceous bedrock formations that will host and enclose the DGR. Such rock mass permeabilities are thought sufficiently low to contribute toward or govern a diffusion-dominated transport regime. To support this concept, hydraulic testing was performed in 2008 and 2009 in two deep boreholes at the proposed repository site, DGR-3 and DGR-4. The hydraulic testing was performed using a straddle-packer tool with a 30.74-m test interval. Sequential tests were performed over the entire open lengths of the boreholes from the F Unit of the Silurian Salina Formation into the Ordovician Gull River Formation, a distance of approximately 635 m. The tests consisted primarily of pressure-pulse tests, with a few slug tests performed in several of the higher permeability Silurian units. The tests are analyzed using the nSIGHTS code, which allows the entire pressure history a test interval has experienced since it was penetrated by the drill bit to be included in the test simulation. nSIGHTS also allows the model fit to the test data to be optimized over an n-dimensional parameter space to ensure that the final solution represents a true global minimum rather than simply a local minimum. The test results show that the Ordovician-age strata above the Coboconk Formation (70+ m below the Cobourg) have average horizontal hydraulic conductivities of 1E-13 m/s or less. Coboconk and Gull River hydraulic conductivities are as high as 1E-11 m

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)

The predictable nature of the Paleozoic sedimentary sequence beneath the Bruce nuclear site in Southern Ontario, Canada

International Nuclear Information System (INIS)

Parmenter, Andrew; Jensen, Mark; Crowe, Richard

2012-01-01

Document available in extended abstract form only. A key aspect of a Deep Geologic Repository (DGR) safety case is the ability to develop and communicate an understanding of the geologic stability and resilience to change at time frames relevant to demonstrating repository performance. As part of an on-going Environmental Assessment, Ontario Power Generation (OPG) recently completed site-specific investigations within an 850 m thick Paleozoic sedimentary sequence beneath the Bruce nuclear site for the proposed development of a DGR for Low and Intermediate Level Waste (L and ILW). As envisioned, the shaft-accessed DGR would be excavated at a nominal depth of 680 m within the low permeability Ordovician argillaceous limestone of the Cobourg Formation, which is overlain by more than 200 m of low permeability Ordovician shale. The geo-scientific investigations revealed a relatively undeformed and laterally continuous architecture within the sedimentary sequence at the repository scale (1.5 km 2 ) and beyond. This paper explores the predictable nature of the sedimentary sequence that has contributed to increasing confidence in an understanding of the spatial distribution of groundwater system properties, deep groundwater system evolution and natural barrier performance. Multi-disciplinary geo-scientific investigations of the Bruce nuclear site were completed in 3 phases between 2006 and 2010. The sub-surface investigations included a deep drilling, coring and in-situ testing program and, the completion of a 19.7 km (9 lines) 2-D seismic reflection survey. The drilling program involved 6 (150 mm dia.) deep boreholes (4-vertical; 2 inclined) that were extended through the sedimentary sequence from 4 drill sites, arranged around the 0.3 km 2 footprint of the proposed repository. The more than 3.8 km of rock core (77 mm dia.) retrieved have provided, in part, a strong basis to understand bedrock lithology and mineralogy, facies assemblages, structure, and oil and gas

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)

Preventive maintenance: optimization of time - based discard decisions at the bruce nuclear generating station

International Nuclear Information System (INIS)

Doyle, E.K.; Jardine, A.K.S.

2001-01-01

The use of various maintenance optimization techniques at Bruce has lead to cost effective preventive maintenance applications for complex systems. As previously reported at ICONE 6 in New Orleans, 1996, several innovative practices reduced Reliability Centered Maintenance costs while maintaining the accuracy of the analysis. The optimization strategy has undergone further evolution and at the present an Integrated Maintenance Program (IMP) is in place where an Expert Panel consisting of all players/experts proceed through each system in a disciplined fashion and reach agreement on all items under a rigorous time frame. It is well known that there are essentially 3 maintenance based actions that can flow from a Maintenance Optimization Analysis: condition based maintenance, time based maintenance and time based discard. The present effort deals with time based discard decisions. Maintenance data from the Remote On-Power Fuel Changing System was used. (author)

Completion of reconstruction for Japan Research Reactor No.3

International Nuclear Information System (INIS)

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

1992-01-01

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

EL3 reactor description and safety analysis report

International Nuclear Information System (INIS)

1969-02-01

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

Course of operators of the RA-3 reactor

International Nuclear Information System (INIS)

Caligiuri, G.A.

1983-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-15

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

AECB staff annual assessment of the Bruce Heavy Water Plant for the year 1994

International Nuclear Information System (INIS)

1995-06-01

This report is the Atomic Energy Control Board staff assessment of the operation of Bruce Heavy Water Plant (BHWP) during 1994. BHWP operation was acceptably safe in 1994. At BHWP, Ontario Hydro did not breach any of the regulations issued under the authority of the Atomic Energy Control Act. There were four minor violations of the BHWP Operating Licence. In all cases, Ontario Hydro exceeded Ontario Hydro government limits for releases to the environment. None of the events threatened public health or the environment. 2 figs

Analysis of effects of calandria tube uncovery under severe accident conditions in CANDU reactors

International Nuclear Information System (INIS)

Rogers, J.T.; Currie, T.C.; Atkinson, J.C.; Dick, R.

1983-01-01

A study is being undertaken for the Atomic Energy Control Board to assess the thermal and hydraulic behaviour of CANDU reactor cores under accident conditions more severe than those normally considered in the licensing process. In this paper, we consider the effects on a coolant channel of the uncovery of a calandria tube by moderator boil-off following a LOCA in a Bruce reactor unit in which emergency cooling is ineffective and the moderator heat sink is impaired by the failure of the moderator cooling system. Calandria tube uncovery and its immediate consequences, as described here, constitute only one part of the entire accident sequence. Other aspects of this sequence as well as results of the analysis of the other accident sequences studied will be described in the final report on the project and in later papers

Control maintenance training program for special safety systems at Bruce B

International Nuclear Information System (INIS)

Reinwald, G.

1997-01-01

It was recognized from the early days of commissioning of Bruce B that Control Maintenance staff would require a level of expertise to be able to maintain Special Safety Systems in proper running order. In the early 80's this was achieved through hands on experience during the original commissioning, troubleshooting and placing of the various systems in service. Control maintenance procedures were developed and implemented as the new systems came available for commissioning, as were operating manuals,training manuals etc. Under the development of the Maintenance Manager, a Conduct of Maintenance section was organized. One of the responsibilities of this section was to develop a series of Maintenance Administrative Procedures (MAPs) that set the standards for maintenance activities including training

AECB staff annual assessment of the Bruce Heavy Water Plant for the year 1995

International Nuclear Information System (INIS)

1996-06-01

The Atomic Energy Control Board's staff annual assessment of the operation of Bruce Heavy Water Plant (BHWP) during 1995. BHWP operation was acceptably safe in 1995. At BHWP, Ontario Hydro complied with the regulations issued under the authority of the Atomic Energy Control Act. AECB is satisfied that BHWP did not pose any undue risk to public health or safety or to the environment. Ontario Hydro met all safety system and safety related system availability targets at BHWP in 1995. The emergency response capability is satisfactory. 2 figs

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

Annual report/1979

International Nuclear Information System (INIS)

1980-04-01

Primary energy demand in Ontario in 1979 was up by 2.9 percent, compared to 2.7 percent in the previous year. Revised forecasts issued in January 1980 indicate Ontario's need for electricity is expected to grow by an average of 3.4 percent annually to the year 2000. Nuclear generation provided 29 percent of the total energy made available by Hydro, and Hydro's eight reactors at Pickering and Bruce continued to rank in the top 36 - four in the top 10 - when compared to the permance of 104 of the world's largest reactors. The provinical legislature's Select Committee on Hydro Affairs examined the safety of the CANDU system and concluded that is is 'acceptably safe'. Faced with reduced forecasts of electrical demands to the year 2000 the Board of Directors decided to stretch out the construction schedule of the Darlington station, to halt construction of the second half of the Bruce Heavy Water Plant D, and to complete but mothball the first half. Construction of Bruce Heavy Water Plant B was completed early in 1979. The A plant produced 599.8 megagrams of reactor-grade heavy water. A control room simulator for Bruce A nuclear generating station was ordered. Agreement was reached on rebuilding faulty boilers at Pickering B. A total of 757.6 megagrams of uranium was used to produce electrical energy and steam. Ontario Hydro continued involvement in uranium exploration. Studies on radioactive waste disposal are being carried out, with emphasis on interim storage and transportation. (LL)

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

International Nuclear Information System (INIS)

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

1992-09-01

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

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

International Nuclear Information System (INIS)

Vicens, Hugo E.; Quintana, Jorge A.

2001-01-01

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

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

International Nuclear Information System (INIS)

Duerigen, S.; Fridman, E.

2011-01-01

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

3D computer visualization and animation of CANDU reactor core

International Nuclear Information System (INIS)

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

1999-01-01

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

Experiment for search for sterile neutrino at SM-3 reactor

Science.gov (United States)

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

2016-11-01

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

Integral test of JENDL-3.3 on fast reactors

International Nuclear Information System (INIS)

Chiba, Gou; Hazama, Taira

2003-05-01

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

Bruce's Magnificent Quartet: Inquiry, Community, Technology and Literacy--Implications for Renewing Qualitative Research in the Twenty-First Century

Science.gov (United States)

Davidson, Judith

2014-01-01

Bruce and Bishop's community informatics work brings forward four critical concepts: inquiry, community, technology, and literacy. These four terms serve as the basis for a discussion of qualitative research in the twenty-first century--what is lacking and what is needed. The author suggests that to resolve the tensions or challenges…

The supply of steam from Candu reactors for heavy water production

International Nuclear Information System (INIS)

Robertson, R.F.S.

1975-09-01

By 1980, Canada's energy needs for D 2 O production will be 420 MW of electrical energy and 3600 MW of thermal energy (as steam). The nature of the process demands that this energy supply be exceptionally stable. Today, production plants are located at or close to nuclear electricity generating sites where advantage can be taken of the low cost of both the electricity and steam produced by nuclear reactors. Reliability of energy supply is achieved by dividing the load between the multiple units which comprise the sites. The present and proposed means of energy supply to the production sites at the Bruce Heavy Water Plant in Ontario and the La Prade Heavy Water Plant in Quebec are described. (author)

Chemical cleaning of the Bruce A steam generators

International Nuclear Information System (INIS)

Le Surf, J.E.; Mason, J.B.; Symmons, W.R.; Yee, F.

1992-01-01

Deposits consisting mostly of oxides and salts and copper metal in the secondary side of the steam generators at the Bruce A Nuclear Generating Station have caused instability in the steam flow and loss of heat capacity, resulting in derating of the units and reduction in power production. Attempts to remove the deposits by pressure pulsing were unsuccessful. Water lancing succeeded in restoring stability, but restrictions on access prevented complete lancing of the tube support plate holes. Chemical cleaning using a modified EPRI-SGOG process has been selected as the best method of removing the deposits. A complete chemical cleaning system has been designed and fabricated for Ontario Hydro by Pacific Nuclear, with support from AECL CANDU and their suppliers. The system consists of self contained modules which are easily interconnected on site. The whole process is controlled from the Control Module, where all parameters are monitored on a computer video screen. The operator can control motorized valves, pumps and heaters from the computer key board. This system incorporates all the advanced technologies and design features that have been developed by Pacific Nuclear in the design, fabrication and operation of many systems for chemical decontamination and cleaning of nuclear systems. 2 figs

75 FR 3217 - J&T Hydro Company; H. Dean Brooks and W. Bruce Cox; Notice of Application for Transfer of License...

Science.gov (United States)

2010-01-20

... DEPARTMENT OF ENERGY Federal Energy Regulatory Commission [Project No. 11392-009] J&T Hydro Company; H. Dean Brooks and W. Bruce Cox; Notice of Application for Transfer of License and Soliciting Comments and Motions To Intervene January 12, 2010. On October 30, 2009, J&T Hydro Company (transferor) and...

AECB staff annual report of Bruce B NGS for the year 1991

International Nuclear Information System (INIS)

1992-11-01

In this account of Bruce NGS B station operation during the year 1991 AECB staff have pointed out non-compliances with the operating licence, which have been few in number and minor in degree of seriousness. There were no exposures of workers to radiation in excess of regulatory limits, but there were contraventions of the ALARA principle. Releases of radioactive material to the environment have been well below the target levels. The performance of the four special safety systems has been good, except for the containment system. A review of the significant event reports and the causes of the events has revealed a lack of a system by which operations and maintenance work could be verified to have been carried out as intended. In operations and maintenance the backlog of work to be done to regularize temporary changes to equipment (removal of jumpers), to carry out preventive maintenance (call-ups), and to make repairs (deficiency reports) has increased from that of the previous year. On the other hand, the station has reduced the number of temporary operating instructions (operating memos) to half of what it was last year. The fretting of steam generator tubes reported last year has not become worse. Nevertheless, inspections continue and modifications to the tube supports are underway. Overall plant chemistry has been acceptable. An Ontario Hydro assessment of the station found the station management's expectations for maintaining the margin of safety in the plant had not been properly communicated to all levels of station staff. The station is now attempting to correct this. Infractions of work protection procedures, which were the subject of many significant events, have led to changes in the procedures and resulted in a major training effort. AECB staff believe that Ontario Hydro has continued to operate Bruce NGS B in a safe manner, but have pointed out areas where improvement is required. (Author)

3D CAD model of the subcritical nuclear reactor of IPN

International Nuclear Information System (INIS)

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

2016-09-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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)

Thermal gradients caused by the CANDU moderator circulation

International Nuclear Information System (INIS)

Mohindra, V.K.; Vartolomei, M.A.; Scharfenberg, R.

2008-01-01

The heavy water moderator circulation system of a CANDU reactor, maintains calandria moderator temperature at power-dependent design values. The temperature differentials between the moderator and the cooler heavy water entering the calandria generate thermal gradients in the reflector and moderator. The resultant small changes in thermal neutron population are detected by the out-of-core ion chambers as small, continuous fluctuations of the Log Rate signals. The impact of the thermal gradients on the frequency of the High Log Rate fluctuations and their amplitude is relatively more pronounced for Bruce A as compared to Bruce B reactors. The root cause of the Log Rate fluctuations was investigated using Bruce Power operating plant information data and the results of the investigation support the interpretation based on the thermal gradient phenomenon. (author)

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)

Mediador cultural ou antropólogo do mal: Bruce Albert e o caso de “A queda do céu”

Directory of Open Access Journals (Sweden)

Karla Alessandra Alves de Souza Ferreira

2017-07-01

Full Text Available Este estudo desenvolve uma análise crítica sobre o fragmento “Postscriptum, quando eu é um outro (e vice-versa”, apresentado na obra A Queda do céu: palavras de um xamã yanomami. O livro foi pensado por um xamã yanomami; Davi Kopenawa, e produzido por um etnólogo francês Bruce Albert. Problematizo o processo de produção do livro onde dois universos culturais se encontram, uma produção literária indígena do povo Yanomami que apresenta uma coautoria. Essa análise busca investigar a postura epistêmica de Bruce Albert como mediador cultural ou “antropólogo do mal” no processo de elaboração do livro, a fim de levantar questionamentos sobre o ato tradutório e suas implicações, destacando os desa os e contribuições apresentadas nesse processo. Nesta direção, esse artigo se fundamenta nos pensamentos da história cultural.

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.

Primary Water Chemistry Control during a Planned Outage at Bruce Power

International Nuclear Information System (INIS)

Ma, Guoping; Nashiem, Rod; Matheson, Shane; Yabar, Berman; Harper, Bill; Roberts, John G.

2012-09-01

Bruce Power has developed a comprehensive outage water chemistry program, which includes both primary and secondary chemistry requirements during planned outages. The purpose of the program is to emphasize the chemistry requirements during outages and subsequent start-ups in order to maintain the integrity of the systems, minimise activity transport and radiation fields, reduce the Carbon-14 release, and to ensure that the requirements are integrated with the outage management program. Prior to a planned outage, Station Chemical Technical Sections identify outage chemistry requirements to Operations and Outage Planning and ensure that work necessary to correct system chemistry issues is within outage work scope. The outage water chemistry program provides direction for establishing alternative sampling locations as demanded by the system configuration during the outage and identifies outage prerequisites for nuclear system purification capabilities. These requirements are contained in an outage checklist. The paper mainly highlights the primary water chemistry issues and chemistry control strategies during planned outages and discusses challenges and successes. (authors)

PR-EDB: Power Reactor Embrittlement Database Version 3

International Nuclear Information System (INIS)

Wang, Jy-An John; Subramani, Ranjit

2008-01-01

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

PR-EDB: Power Reactor Embrittlement Database - Version 3

Energy Technology Data Exchange (ETDEWEB)

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

2008-03-01

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

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

International Nuclear Information System (INIS)

1983-04-01

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

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Feasibility analysis of the utilization of moderator heat for agricultural and aquacultural purposes, Bruce nuclear power development

International Nuclear Information System (INIS)

1977-12-01

A study is presented of the feasibility of using moderator reject heat from the Bruce nuclear power development either to heat greenhouses or to aid in a warm water hatchery or aquaculture operation. The study examines heat extraction and delivery plans, reliability of supply, pricing schedules, the Ontario greenhouse industry, site selection criteria, water transmission and distribution, costs, approvals required, and a construction timetable. Total system analysis shows that a greenhouse facility would be viable but the aquaculture/hatchery scheme is more cost-effective. (E.C.B.)

Benchmark tests of JENDL-3.2 for thermal and fast reactors

International Nuclear Information System (INIS)

Takano, Hideki

1995-01-01

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

Nuclear research reactor 0.5 to 3 MW

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-05-15

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

Nuclear research reactor 0.5 to 3 MW

International Nuclear Information System (INIS)

1992-05-01

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

A progress review of Ontario Hydro's nuclear generation and heavy water production programs

International Nuclear Information System (INIS)

Kee, F.J.; Woodhead, L.W.

Performance and economics of CANDU reactors in service are described. Progress of commissioning, construction and planning of reactors at Pickering, Bruce, and Darlington is outlined. Heavy water production is reviewed. (E.C.B.)

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

Dolomitized bryozoan bioherms from the Lower Silurian Manitoulin Formation, Bruce Peninsula, Ontario

Energy Technology Data Exchange (ETDEWEB)

Anastas, A S; Coniglo, M [Waterloo Univ., ON (Canada)

1992-06-01

Several small, previously undescribed bioherms are present in the shallow shelf dolostones of the Manitoulin Formation at the Cabot Head and Wingfield Basin localities in the northernmost portion of the Bruce Peninsula region of southern Ontario. The bioherms, commonly associated with carbonate tempestites, range from 0.3 to 1.0 m in height and 0.9 to 2.5 m in width and are composed of bafflestones-floatstones and minor bindstones. The chief components of the bioherms are dolomitized lime mud and branching bryozoans. Bioherm building by bryozoans, although common in the ancient record, represents a great divergence from the mostly accessory frame encrusting role of bryozoans in modern environments. Minor skeletal components of the bioherms include echinoderms, rugose and tabulate corals and brachiopods. Laminar encrusting bryozoans exist in the top 10 cm of one of the bioherms. Some of the bioherms show evidence of water agitation that may be the result of current action induced by storm or tidal processes. The occurrence of the bioherms stretches the already known Llandoverian reef complex on Manitoulin Island further to the south. The reason why these bioherms did not reach sizes comparable to large Llandoverian or Wenlockian reefs and did not make the shift to coral-stromaporoid community is probably related to a complex interaction of factors such as community development, bathymetry, clasticity and salinity. 41 refs., 4 figs.

Bruce NGS B risk assessment (BBRA) peer review process

International Nuclear Information System (INIS)

Kaasalainen, S.; Crocker, W.P.; Webb, W.A.

2001-01-01

Risk-informed decision making is considered an effective approach to managing the risk of nuclear power plant operation in a competitive market. Hence, increased reliance on the station probabilistic risk assessments (PRAs) to provide risk perspective inputs is inevitable. With increased reliance on the PRAs it is imperative that PRAs have the characteristics necessary to provide the required information. Recognizing the increased requirements on nuclear power plant PRAs the nuclear industry in the United States has expended significant effort over the past few years defining the required characteristics of a PRA for various applications. More recently several owners groups have drafted guidelines for PRA certification and several U.S. utilities have had their PRAs certified. During the year 2000 Ontario Power Generation, Nuclear (OPG,N) subjected the PRA of one of its stations to the U.S. style certification process. The PRA selected for this process was the Bruce B Risk Assessment (BBRA). BBRA was chosen for this process since it is the first OPG, N PRA to be used for risk-informed applications. However, the strengths of the BBRA identified from the certification process and the lessons learned are also largely applicable to the other OPG, N plant PRAs due to the use of similar methods and tools

Use of plate fuel elements for the RA3 reactor

International Nuclear Information System (INIS)

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

1992-01-01

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

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

International Nuclear Information System (INIS)

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

1993-11-01

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

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

International Nuclear Information System (INIS)

Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi

2000-06-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-06-01

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

Leadership Actions to Improve Nuclear Safety Culture

International Nuclear Information System (INIS)

Clewett, L.K.

2016-01-01

The challenge many leaders face is how to effectively implement and then utilise the results of Safety Culture surveys. Bruce Power has recently successfully implemented changes to the Safety Culture survey process including how corrective actions were identified and implemented. The actions taken in response to the latest survey have proven effective with step change performance noted. Nuclear Safety is a core value for Bruce Power. Nuclear Safety at Bruce Power is based on the following four pillars: reactor safety, industrial safety, radiological safety and environmental safety. Processes and practices are in place to achieve a healthy Nuclear Safety Culture within Bruce Power such that nuclear safety is the overriding priority. This governance is based on industry leading practices which monitor, asses and take action to drive continual improvements in the Nuclear Safety Culture within Bruce Power.

AECB staff annual assessment of the Bruce A Nuclear Generating Station for the year 1994

International Nuclear Information System (INIS)

1995-06-01

AECB believe that Ontario Hydro operated Bruce A in a safe manner during 1994, and that the risk to workers and the public has been maintained at an acceptably low level. Radiation doses to workers and releases to the environment were well below regulatory limits. All special safety systems met availability targets. We noted improvements in operation and maintenance but some further improvements are still required. This is particularly true of the station's compliance with the Operating Licence. AECB believe that the station continues to be well managed, with a high priority placed on safety. However, there is a need for increased capability in the area of safety analysis and assessment. 4 tabs., 4 figs

Retrofitting alarm prioritization at Bruce A: strategy development and implementation experience

International Nuclear Information System (INIS)

Davey, E.; Hickey, D.; Babcock, B.

1997-01-01

A prioritization strategy for computer-displayed control room alarms has been developed for Bruce A to better assist operations staff in visually identifying key alarms and judging the relative importance of alarms. The strategy consists of assigning each alarm indicative of a problem to be addressed to one of five priority categories. Each alarm is assigned to an alarm category based on an off-line analysis of the consequence and response characteristics applicable to the alarm for three plant operating contexts. The colour of the alarm message is used to convey the priority category of each alarm in computer-based alarm displays. In addition, alarms indicative of non-problematic changes in the state of plant equipment and processes are given a separate colour assignment to visually differentiate them from alarms indicative of problems. This paper outlines the user-based approach employed in the prioritization strategy development, describes the key features of the prioritization strategy adopted, and discusses the initial experience in systematically determining the priority assignments for all 6000 computer-based alarms associated with each generating unit. (author)

Floral micromorphology of the genus Ensete Bruce ex Horan. (Musaceae in Thailand

Directory of Open Access Journals (Sweden)

Wandee Inta

2015-09-01

Full Text Available To fulfil scarce and incomplete information on floral micromorphology of ensets (Ensete Bruce ex Horan. in the banana family (Musaceae, a comparative anatomical study of two species: E. glaucum (Roxb. Cheesman and E. superbum (Roxb. Cheesman, native to Thailand was conducted. It was found that, apart from five fertile stamens presented in other members of the Musaceae family, both ensets possess a short staminode. It is suggested from this investigation that six is the basic number of Ensete androecial whorl and the taxa could secure the most primitive status within the family and the Zingiberales order, of which stamen numbers are reduced. The results also indicated that the vascular bundle position in compound tepal, the vascular patterns in vascular zone of ovary and cell shapes of stigma epidermis and the ovary cortex are of systematic significance in conjunction with pollen size and exine ornamentation. These useful micromorphological characters can be further applied for identification of other Ensete species distribute elsewhere in the world.

Integral test of JENDL-3.3 for thermal reactors

International Nuclear Information System (INIS)

Okumura, Keisuke; Mori, Takamasa

2003-01-01

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

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

International Nuclear Information System (INIS)

Shi, Dunfu

2015-01-01

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

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

CERN Document Server

Chiba, G

2002-01-01

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

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

International Nuclear Information System (INIS)

Rosli Darmawan; Mariah Adam

2012-01-01

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

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

International Nuclear Information System (INIS)

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

1983-07-01

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

RA reactor exploitation, task 3.08/01

International Nuclear Information System (INIS)

Zecevic, V.

1963-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1993-11-01

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

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

International Nuclear Information System (INIS)

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

1982-03-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-04-01

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

Stress corrosion cracking experience in steam generators at Bruce NGS

International Nuclear Information System (INIS)

King, P.J.; Gonzalez, F.; Brown, J.

1993-01-01

In late 1990 and through 1991, units 1 and 2 at the Bruce A Nuclear Generating Station (BNGS-A) experienced a number of steam generator tube leaks. Tube failures were identified by eddy current to be circumferential cracks at U-bend supports on the hot-leg side of the boilers. In late 1991, tubes were removed from these units for failure characterization. Two active failure modes were found: corrosion fatigue in both units 1 and 2 and stress corrosion cracking (SCC) in unit 2. In unit 2, lead was found in deposits, on tubes, and in cracks, and the cracking was mixed-mode: transgranular and intergranular. This convincingly indicated the involvement of lead in the stress corrosion cracking failures. A program of inspection and tube removals was carried out to investigate more fully the extent of the problem. This program found significant cracking only in lead-affected boilers in unit 2, and also revealed a limited extent of non-lead-related intergranular stress corrosion cracking in other boilers and units. Various aspects of the failures and tube examinations are presented in this paper. Included is discussion of the cracking morphology, measured crack size distributions, and chemical analysis of tube surfaces, crack faces, and deposits -- with particular emphasis on lead

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.

Optimization and implementation study of plutonium disposition using existing CANDU Reactors. Final report

International Nuclear Information System (INIS)

1996-09-01

Since early 1994, the Department of Energy has been sponsoring studies aimed at evaluating the merits of disposing of surplus US weapons plutonium as Mixed Oxide (MOX) fuel in existing commercial Canadian Pressurized Heavy Water reactors, known as CANDU's. The first report, submitted to DOE in July, 1994 (the 1994 Executive Summary is attached), identified practical and safe options for the consumption of 50 to 100 tons of plutonium in 25 years in some of the existing CANDU reactors operating the Bruce A generating station, on Lake Huron, about 300 km north east of Detroit. By designing the fuel and nuclear performance to operate within existing experience and operating/performance envelope, and by utilizing existing fuel fabrication and transportation facilities and methods, a low cost, low risk method for long term plutonium disposition was developed. In December, 1995, in response to evolving Mission Requirements, the DOE requested a further study of the CANDU option with emphasis on more rapid disposition of the plutonium, and retaining the early start and low risk features of the earlier work. This report is the result of that additional work

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.)

Exocrine Gland-Secreting Peptide 1 Is a Key Chemosensory Signal Responsible for the Bruce Effect in Mice.

Science.gov (United States)

Hattori, Tatsuya; Osakada, Takuya; Masaoka, Takuto; Ooyama, Rumi; Horio, Nao; Mogi, Kazutaka; Nagasawa, Miho; Haga-Yamanaka, Sachiko; Touhara, Kazushige; Kikusui, Takefumi

2017-10-23

The Bruce effect refers to pregnancy termination in recently pregnant female rodents upon exposure to unfamiliar males [1]. This event occurs in specific combinations of laboratory mouse strains via the vomeronasal system [2, 3]; however, the responsible chemosensory signals have not been fully identified. Here we demonstrate that the male pheromone exocrine gland-secreting peptide 1 (ESP1) is one of the key factors that causes pregnancy block. Female mice exhibited high pregnancy failure rates upon encountering males that secreted different levels of ESP1 compared to the mated male. The effect was not observed in mice that lacked the ESP1 receptor, V2Rp5, which is expressed in vomeronasal sensory neurons. Prolactin surges in the blood after mating, which are essential for maintaining luteal function, were suppressed by ESP1 exposure, suggesting that a neuroendocrine mechanism underlies ESP1-mediated pregnancy failure. The single peptide pheromone ESP1 conveys not only maleness to promote female receptivity but also the males' characteristics to facilitate memorization of the mating partner. Copyright © 2017 Elsevier Ltd. All rights reserved.

Ignition access in a D-3He helical reactor

International Nuclear Information System (INIS)

Mitarai, Osamu

2003-01-01

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

Experience in ultrasonic gap measurement between calandria tubes and liquid injection shutdown systems nozzles in Bruce Nuclear Generating Station

International Nuclear Information System (INIS)

Abucay, R.C.; Mahil, K.S.; Goszczynski, J.J.

1995-01-01

The gaps between calandria tubes (CT) and Liquid Injection Shutdown System (LISS) nozzles at the Bruce Nuclear Generating Station ''A'' (Bruce A) are known to decrease with time due to radiation induced creep/sag of the calandria tubes. If this gap decreases to a point where the calandria tubes come into contact with the LISS nozzle, the calandria tubes could fail as a result of fretting damage. Proximity measurements were needed to verify the analytical models and ensure that CT/LISS nozzle contact does not occur earlier than predicted. The technique used was originally developed at Ontario Hydro Technologies (formerly Ontario Hydro Research Division) in the late seventies and put into practical use by Research and Productivity Council (RPC) of New Brunswick, who carried out similar measurements at Point Lepreau NGS in 1989 and 1991. The gap measurement was accomplished y inserting an inspection probe, containing four ultrasonic transducers (2 to measure gaps and 2 to check for probe tilt) and a Fredericks electrolytic potentiometer as a probe rotational sensor, inside LISS Nozzle number-sign 7. The ultrasonic measurements were fed to a system computer that was programmed to convert the readings into fully compensated gaps, taking into account moderator heavy water temperature and probe tilt. Since the measured gaps were found to be generally larger than predicted, the time to CT/LISS nozzle contact is now being re-evaluated and the planned LISS nozzle replacement will likely be deferred, resulting in considerable savings

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

International Nuclear Information System (INIS)

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

1977-01-01

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

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

International Nuclear Information System (INIS)

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

1977-01-01

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

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

International Nuclear Information System (INIS)

Kaliatka, Algirdas; Bubelis, Evaldas; Uspuras, Eugenijus

2003-01-01

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

My City of Ruins: Bruce Springsteen e l’utopia fra le rovine

Directory of Open Access Journals (Sweden)

Enrico Botta

2011-09-01

Full Text Available The paper focuses on My City of Ruins, the song that Bruce Springsteen sang—ten days after the 09/11 terrorist attacks—for "America: A Tribute to Heroes" and that was released in the concept album The Rising in 2002. The essay aims to highlight how the song describes the post 09/11 New York City by opposing the themes of “ruins” and “utopia.” From a textual and musical point of view, My City of Ruins is, in fact, composed of a double structure that balances different feelings: the fear, pain, and loneliness of the victims, in the rock-blues first section; the faith, love, and hope, in the gospel second part. Furthermore, the paper tries to point out how My City of Ruins no longer describes the symbolic ruins of a foreign past—in line with the nineteenth and twentieth-century American cultural tradition of the Grand Tour—but defines the physical signs of a definitively collapsed “American dream,” which can survive only in a utopian and spiritual “Promised Land.”

Removal of bulk contaminants from radioactive waste water at Bruce A using a clay based flocculent system

International Nuclear Information System (INIS)

Davloor, R.; Harper, B.

2011-01-01

Bruce Power's Bruce Nuclear Generating Station 'A', located on Lake Huron, has a treatment system that processes all aqueous radioactive waste water originating from the station. This Active Liquid Waste Treatment System (ALWTS) consists of collection tanks for the collection of radioactive waste water, a Pre-Treatment System (PTS) for the removal of bulk contaminants and suspended solids, a Reverse Osmosis System (ROS) to remove dissolved solids, an Evaporation and Solidification System (ESS) to concentrate and immobilize solids contained in concentrated waste streams from the ROS, and discharge tanks for the dispersal of the treated water. The ALWTS has been in continuous service since 1999 and is used to treat approximately 100,000 litres of Active Liquid Waste (ALW) each day. With the exception of tritium, it discharges waste water containing near zero concentrations of radioactive and conventional contaminants to the lake. The original design of the Bruce A ALWTS used a Backwashable Filtration System (BFS) to provide solids free water to the ROS, as measured by the Silt Density Index (SDI). During commissioning, the BFS was not successful in backwashing the solids from the filter elements. For approximately one year, a temporary solution was implemented using a Disposable Filtration System (DFS). A cationic polymer was added upstream of the DFS to agglomerate the solids. The system proved to be highly unreliable. It was difficult to agglomerate solids in the waste stream containing high amounts of detergent. As a result, DFS consumption was high and very costly. The SDI specification for the RO membrane was not always met, resulting in a quick decline of performance of the first stage ROS membranes in the treatment process. In addition, the excess cationic polymer in the RO feed caused the membranes to become fouled. In-house station staff, together with personnel from Colloid Environmental Technologies (CETCO) Company, worked to develop and

Removal of bulk contaminants from radioactive waste water at Bruce A using a clay based flocculent system

Energy Technology Data Exchange (ETDEWEB)

Davloor, R.; Harper, B. [Bruce Power, Tiverton, ON (Canada)

2011-07-01

Bruce Power's Bruce Nuclear Generating Station 'A', located on Lake Huron, has a treatment system that processes all aqueous radioactive waste water originating from the station. This Active Liquid Waste Treatment System (ALWTS) consists of collection tanks for the collection of radioactive waste water, a Pre-Treatment System (PTS) for the removal of bulk contaminants and suspended solids, a Reverse Osmosis System (ROS) to remove dissolved solids, an Evaporation and Solidification System (ESS) to concentrate and immobilize solids contained in concentrated waste streams from the ROS, and discharge tanks for the dispersal of the treated water. The ALWTS has been in continuous service since 1999 and is used to treat approximately 100,000 litres of Active Liquid Waste (ALW) each day. With the exception of tritium, it discharges waste water containing near zero concentrations of radioactive and conventional contaminants to the lake. The original design of the Bruce A ALWTS used a Backwashable Filtration System (BFS) to provide solids free water to the ROS, as measured by the Silt Density Index (SDI). During commissioning, the BFS was not successful in backwashing the solids from the filter elements. For approximately one year, a temporary solution was implemented using a Disposable Filtration System (DFS). A cationic polymer was added upstream of the DFS to agglomerate the solids. The system proved to be highly unreliable. It was difficult to agglomerate solids in the waste stream containing high amounts of detergent. As a result, DFS consumption was high and very costly. The SDI specification for the RO membrane was not always met, resulting in a quick decline of performance of the first stage ROS membranes in the treatment process. In addition, the excess cationic polymer in the RO feed caused the membranes to become fouled. In-house station staff, together with personnel from Colloid Environmental Technologies (CETCO) Company, worked to develop and

GRIMH3: A new reactor calculation code at Savannah River Site

International Nuclear Information System (INIS)

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

1993-01-01

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

Annual report, 1978

International Nuclear Information System (INIS)

1979-04-01

In 1978, for the first time, nuclear generators supplied more electricity than coal-fired units: 30 percent of the total compared with 28 percent for coal. Energy demand in Ontario was up by 2.7 percent. No new commitments for generating stations were made, and work on committed stations was to be slowed until the generation expansion program had been fully reviewed. Atomic Energy of Canada Ltd. and Ontario Hydro have agreed to develop a nuclear wastes demonstration facility. The Select Committee of the Ontario Legislature on Ontario Hydro affairs investigated the costs of the Bruce heavy water plants and found no evidence of mismanagement. The Royal Commission on Electric Power Planning issued an interim report on nuclear power which recognized the need for and safety of the CANDU system. Reactors at Pickering and Bruce achieved an overall capacity factor of 81 percent. The third Bruce A unit was started up. Work on Bruce B and Pickering B was well underway. Bruce Heavy Water Plant B was virtually complete, but work was stopped on the second half of the Bruce D plant. Plans for the first half of Bruce D will be reviewed. Site preparation and excavation continued for the Darlington generating station. (LL)

Explorability and predictability of the paleozoic sedimentary sequence beneath the Bruce nuclear site

International Nuclear Information System (INIS)

Parmenter, A.; Jensen, M.; Crowe, R.; Raven, K.

2011-01-01

Ontario Power Generation (OPG) is proposing to develop a Deep Geologic Repository (DGR) for the long-term management of its Low and Intermediate Level Waste (L&ILW) at the Bruce nuclear site located in the Municipality of Kincardine, Ontario. A 4-year program of geoscientific studies to assess the suitability of the 850 m thick Palaeozoic age sedimentary sequence beneath the site to host the DGR was completed in 2010. The studies provide evidence of a geologic setting in which the DGR concept would be safely implemented at a nominal depth of 680 m within the argillaceous limestone of the Cobourg Formation. This paper describes the geologic framework of the Bruce nuclear site with a focus on illustrating the high degree of stratigraphic continuity and traceability at site-specific and regional scales within the Ordovician sediments proposed to host and enclose the DGR. As part of the site-specific studies, a program of deep drilling/coring (6 boreholes) and in-situ testing through the sedimentary sequence was completed from 4 drill sites situated beyond the DGR footprint, approximately 1 km apart. Core logging reveals that the stratigraphic sequence comprises 34 distinct bedrock formations/members/units consistent with the known regional stratigraphic framework. These layered sedimentary formations dip 0.6 o (~10 m/km) to the southwest with highly uniform thicknesses both at the site- and regional-scale, particularly, the Ordovician sediments, which vary on the order of metres. The occurrence of steeply-dipping faults within the sedimentary sequence is not revealed through surface outcrop fracture mapping, micro-seismic (M ≥ 1) monitoring, inclined borehole coring or intersection of hydrothermal type dolomitized reservoir systems. Potential fault structures, interpreted from a 2-D seismic survey, were targeted by angled boreholes which found no evidence for their existence. Formation specific continuity is also evidence by the lateral traceability of physical rock

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)

White spots on Smoke rings by Bruce Nauman: a case study on contemporary art conservation using microanalytical techniques.

Science.gov (United States)

Mafalda, Ana Cardeira; da Câmara, Rodrigo Bettencourt; Strzelec, Patrick; Schiavon, Nick; Mirão, José; Candeias, António; Carvalho, Maria Luísa; Manso, Marta

2015-02-01

The artwork "Smoke Rings: Two Concentric Tunnels, Non-Communicating" by Bruce Nauman represents a case study of corrosion of a black patina-coated Al-alloy contemporary artwork. The main concern over this artwork was the widespread presence of white spots on its surface. Alloy substrate, patina, and white spots were characterized by means of energy-dispersive X-ray fluorescence and scanning electron microscopy with energy-dispersive spectroscopy. Alloy substrate was identified as an aluminum alloy 6,000 series Al-Si-Mg. Patina's identified composition confirmed the documentation provided by the atelier. Concerning the white spots, zircon particles were found on patina surface as external elements.

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-10-01

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

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

International Nuclear Information System (INIS)

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

2012-01-01

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

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

International Nuclear Information System (INIS)

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

1992-01-01

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

The Flamanville 3 EPR reactor; Le reacteur EPR Flamanville 3

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

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

Reversed field pinch reactor study 3

International Nuclear Information System (INIS)

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

1977-12-01

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

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

International Nuclear Information System (INIS)

Anon.

1976-01-01

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

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

International Nuclear Information System (INIS)

1979-11-01

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

Development of telerobotic systems for reactor decommissioning, (3)

International Nuclear Information System (INIS)

Usui, Hozumi; Fujii, Yoshio; Shinohara, Yoshikuni

1991-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-11-01

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

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

International Nuclear Information System (INIS)

El-Sahlamy, N.M.

2017-01-01

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

The qualification of U3O8 as research reactor fuel

International Nuclear Information System (INIS)

Krull, W.

1983-01-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

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

International Nuclear Information System (INIS)

Khan, Jahirul Haque

2013-01-01

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

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

International Nuclear Information System (INIS)

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

1974-01-01

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

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

International Nuclear Information System (INIS)

Raynal, A.

1987-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1969-02-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1969-02-01

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

Uranium-fuel thermal reactor benchmark testing of CENDL-3

International Nuclear Information System (INIS)

Liu Ping

2001-01-01

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

Investigations related to a one-piece removal of the reactor block in the frame of the JRR-3 reconstruction program

International Nuclear Information System (INIS)

Onishi, N.; Kanenari, A.; Futamura, Y.; Sakurai, H.; Suzuki, S.; Nagase, T.; Iwatani, A.; Otsubo, F.

1987-01-01

In the Japan Atomic Energy Research Institute (JAERI), an outdated research reactor (Japan Research Reactor No.3; JRR-3) was removed to a storage facility between October 14th and November 7th, 1986. The removal of the 2250-ton reactor block (10 x 10 x 10 m) was performed as a part of a program to replace the JRR-3's core (10-MW thermal) with an upgraded research reactor core. The heavy water and fuel elements were taken out from the JRR-3 before removal work began. The reactor block was raised about 3.7 meters, using a 12-cubic meter steel frame and a center-hole jack system. The reactor block was then transported horizontally about 34 meters on steel rails, using four 100-ton jacks, to a storage facility. Finally, the reactor block was lowered 14 meters into the storage facility. After the reactor block was stored, a new 20-MW thermal, light-water moderated and cooled JRR-3 core will be built, with criticality targeted for 1989

Decommissioning of the BR3 pressurized-water reactor

International Nuclear Information System (INIS)

Massaut, V.

1996-01-01

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

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

International Nuclear Information System (INIS)

Fernandes, Gustavo Henrique Nazareno

2018-01-01

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

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

International Nuclear Information System (INIS)

1966-01-01

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

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

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

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

International Nuclear Information System (INIS)

Edwards, Robert; Huang, Zhengyu

2001-01-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

Advances in Reactor physics, mathematics and computation. Volume 3

Energy Technology Data Exchange (ETDEWEB)

1987-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

Preparations for the shipment of RA-3 reactor irradiated fuel

International Nuclear Information System (INIS)

Goldschmidt, Adrian; Novara, Oscar; Lafuente, Jose

2002-01-01

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

Identity as Fleeting Fashion? Revealing the Background of Popularity Through Bosniaherzegovina’s Adoption of Bruce Lee

Directory of Open Access Journals (Sweden)

Alenka Bartulović

2007-12-01

Full Text Available This paper reflects on the background to popular research on identification processes in modern ethnology and cultural or social anthropology. By discussing themes closely connected to identity (national and popular culture, and the construction of Others, the author reveals that identity is not merely a redundant trend that researchers recklessly pursue, but that it reflects the need to reveal existing uncertainties and is an attempt to revise errors made to date, as well as lapses of the discipline we work in. The author primarily focuses on national identity and its situational transformations and intertwining with other existing identities. Bosnia-Herzegovina is placed at the center of interest. This theoretically oriented text concludes by analyzing the idea of erecting a monument to Bruce Lee in Mostar, which reveals the full complexity of identity processes around a world marked by intense and unique global and transnational currents.

Monitoring the risk of loss of heat sink during plant shutdowns at Bruce Generating Station 'A'

International Nuclear Information System (INIS)

Krishnan, K.S.; Mancuso, F.; Vecchiarelli, D.

1996-01-01

A relatively simple loss of shutdown heat sink fault tree model has been developed and used during unit outages at Bruce Nuclear Generation Station 'A' to assess, from a risk and reliability perspective, alternative heat sink strategies and to aid in decisions on allowable outage configurations. The model is adjusted to reflect the various unit configurations planned during a specific outage, and identifies events and event combinations leading to loss of fuel cooling. The calculated failure frequencies are compared to the limits consistent with corporate and international public safety goals. The importance measures generated by the interrogation of the fault tree model for each outage configuration are also used to reschedule configurations with high fuel damage frequency later into the outage and to control the configurations with relatively high probability of fuel damage to short intervals at the most appropriate time into the outage. (author)

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

International Nuclear Information System (INIS)

Gonzalez M, J.L.

1994-01-01

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

Cobalt-60 production in CANDU power reactors

International Nuclear Information System (INIS)

Slack, J.; Norton, J.L.; Malkoske, G.R.

2003-01-01

therapy machines. Today the majority of the cancer therapy cobalt-60 sources used in the world are manufactured using material from the NRU reactor in Chalk River. The same technology that was used for producing cobalt-60 in a research reactor was then adapted and transferred for use in a CANDU power reactor. In the early 1970s, in co-operation with Ontario Power Generation (formerly Ontario Hydro), bulk cobalt-60 production was initiated in the four Pickering A CANDU reactors located east of Toronto. This was the first full scale production of millions of curies of cobalt-60 per year. As the demand and acceptance of sterilization of medical products grew, MDS Nordion expanded its bulk supply by installing the proprietary Canadian technology in additional CANDUs. Over the years MDS Nordion has partnered with CANDU reactor owners to produce cobalt-60 at various sites. CANDU reactors that have, or are still producing cobalt-60, include Pickering A, Pickering B, Gentilly 2, Embalse in Argentina, and Bruce B. In conclusion, the technology for cobalt-60 production in CANDU reactors, designed and developed by MDS Nordion and Atomic Energy of Canada, has been safely, economically and successfully employed in CANDU reactors with over 195 reactor years of production. Today over forty percent of the world's disposable medical supplies are made safer through sterilization using cobalt-60 sources from MDS Nordion. Over the past 40 years, MDS Nordion with its CANDU reactor owner partners, has safely and reliably shipped more than 500 million curies of cobalt-60 sources to customers around the world. MDS Nordion is presently adding three more CANDU power reactors to its supply chain. These three additional cobalt producing CANDU's will help supplement the ability of the health care industry to provide safe, sterile, medical disposable products to people around the world. As new applications for cobalt-60 are identified, and the demand for bulk cobalt-60 increases, MDS Nordion and AECL

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

International Nuclear Information System (INIS)

Merk, B.; Kliem, S.

2013-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-06-15

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1963-12-15

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

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

International Nuclear Information System (INIS)

1983-07-01

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

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

International Nuclear Information System (INIS)

Biswas, J.K.

1992-01-01

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

TU Electric reactor physics model verification: Power reactor benchmark

International Nuclear Information System (INIS)

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

1988-01-01

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

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

Science.gov (United States)

Longoni, Gianluca; Anderson, Stanwood L.

2009-08-01

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

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

International Nuclear Information System (INIS)

Xia, Hong; Li, Bin; Liu, Jianxin

2014-01-01

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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

3. International conference on catalysis in membrane reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-09-01

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

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

International Nuclear Information System (INIS)

Longoni, G.

2010-01-01

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

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

International Nuclear Information System (INIS)

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

1996-06-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-06-01

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

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

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

The dew point response of the annulus gas system of Bruce NGS A

International Nuclear Information System (INIS)

Kenchington, J.; Ellis, P.J.; Meranda, D.

1983-01-01

The dew point response of the Annulus Gas System in Bruce A, Units 1 and 2 has been modelled in order to alert the operator of the presence of heavy water and to estimate the leak rate into the annulus. The computer model can be easily adapted to determine the Annulus Gas System dew point response in any station. It models the complex arrangement of the system and the transportation of moisture through the annuli by a combination of plug flow and mixing of CO 2 and D 2 O vapor. It predicts the response of the dew point monitor for a range of specified leak rates and positions of a leaking channel in a string of channels. This model has been used to calculate the variation of dew point and rate of change of dew point with respect to time (t). It shows that there is a maximum in the rate of dew point change (dT/dt) with respect to the corresponding dew point (T). This maximum is unique for a given leak rate and channel position. It is independent of the starting time for the leak. The computer programme has been verified by an analytical solution for the model

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

Energy Technology Data Exchange (ETDEWEB)

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

2003-04-01

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Research reactors in Argentina

International Nuclear Information System (INIS)

Carlos Ruben Calabrese

1999-01-01

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

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

International Nuclear Information System (INIS)

Chi-Mei Lee; Kuo-Tsang Hung

2006-01-01

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

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

International Nuclear Information System (INIS)

1992-04-01

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

The ARIES-III D-3He tokamak reactor

International Nuclear Information System (INIS)

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

1992-01-01

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

Annealing of the BR3 reactor pressure vessel

International Nuclear Information System (INIS)

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

1985-01-01

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

RB Research nuclear reactor RB reactor, Annual report for 2000

International Nuclear Information System (INIS)

Milosevic, M.

2000-12-01

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

Reactor science and technology: operation and control of reactors

International Nuclear Information System (INIS)

Qiu Junlong

1994-01-01

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

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

International Nuclear Information System (INIS)

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

2002-03-01

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

3 D flow computations under a reactor vessel closure head

International Nuclear Information System (INIS)

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

1995-12-01

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

Review on the seismic safety of JRR-3 according to the revised regulatory code on seismic design for nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Kobayashi, Tetsuya; Araki, Masaaki; Ohba, Toshinobu; Torii, Yoshiya [Japan Atomic Energy Agency, Tokai, Ibaraki (Japan); Takeuchi, Masaki [Nuclear Safety Commission (Japan)

2012-03-15

JRR-3(Japan Research Reactor No.3) with the thermal power of 20MW is a light water moderated and cooled, swimming pool type research reactor. JRR-3 has been operated without major troubles. This paper presents about review on the seismic safety of JRR-3 according to the revised regulatory code on seismic design for nuclear reactors. In addition, some topics concerning damages in JRR-3 due to the Great East Japan Earthquake are presented. (author)

The case for new nuclear

International Nuclear Information System (INIS)

Luxat, J.C.

2013-01-01

Over a 22 year period from 1971 to 1993 a total of 20 reactor units were brought into service - an average of approximately one unit per year. Ontario Hydro constructed the four-unit Pickering A station, four units at Bruce A, four units at Pickering B, four units at Bruce B and four units at Darlington during this period. This represents a capacity of nearly 14,000 MW, as shown in Figure 1. During this period there was a large increase in industrial capacity in Ontario, particularly in manufacturing, driven in large measure by the incentives offered by low electricity prices, skilled workers and a good health care system. Subsequently in the mid-1990's the Pickering A and Bruce A units were laid up and maintenance efforts were focused on the Pickering B, Bruce B and Darlington stations. Two of the four units at Pickering A were returned to service in the early 2000's and the four units of Bruce A were returned to service with two units being refurbished. By 2010 nuclear capacity in the province had returned to 12,800 MW. The Ontario Long Term Energy Plan (LTEP) announced at the beginning of December does not include new build nuclear but does include refurbishment of the Darlington station as well as two units at Bruce A and four units at Bruce B. The six units at Pickering will be shut down by 2020. As shown in Figure 1, this will reduce the nuclear capacity from the current 12,800 MW to 8000 MW when the Pickering A and B units are removed from service in 2020 and the refurbishment of Darlington and Bruce units proceeds starting in 2016 and projected to complete by 2031. This will be the lowest nuclear generating capacity in the province since 1985. (author)

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

International Nuclear Information System (INIS)

Koch, R.; Findeisen, A.

1986-04-01

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

Tight fitting garter springs-MODAR

Energy Technology Data Exchange (ETDEWEB)

Kazimer, D. [Bruce Power, Tiverton, Ontario (Canada)

2011-07-01

Annulus spacers are used in CANDU reactors to maintain the annular gap between two tubes - an inner pressure tube (PT) and the outer calandria tube (CT). Typically four annulus spacers are used in one fuel channel assembly, each at a specified axial position. Bruce Unit 8 and many other CANDU units were constructed with tight-fitting garter springs (TFGS). The TFGS were not designed to be detected or relocated by the conventional tool, Spacer Location And Repositioning (SLAR) processes. Due to non-optimal 'As Left' construction locations for the Bruce Unit 8 TFGS, PT/CT contact has been predicted to occur well prior to its End of Life (EOL). Bruce Power entered a Project with AECL-CRL to design, manufacture and test and implement a new tooling system that would detect and reposition tight fitting annulus spacers. (author)

Canada country report

International Nuclear Information System (INIS)

Cottrill, Cheryl

2008-01-01

1 - Nuclear 2007 highlights: New Build Applications and Environmental Assessments (Ontario Power Generation (OPG), Bruce Power, Bruce Power Alberta), Refurbishments (Bruce Power's Bruce A Units 1 and 2 Restart Project, NB Power's Refurbishment of Point Lepreau, New Brunswick, Atomic Energy of Canada Limited (AECL) NRU 50. Anniversary, expansion of the solid radioactive waste storage facilities at Gentilly-2 nuclear generating station, Ontario Power Generation (OPG) Deep Geologic Repository..); 2. Nuclear overview: a. Energy policy (Future of nuclear power, state of the projects, schedule, Refurbishment), b. Public acceptance, Statements from Government Officials in Canada; c. Nuclear equipment (number and type); d. Nuclear waste management, Deep Geologic Repository; e. Nuclear research at AECL; f. Other nuclear activities (Cameco Corporation, MDS Nordion); 3. Nuclear competencies; 4. WIN 2007 Main Achievements: GIRLS Science Club, Skills Canada, WiN-Canada Web site, Book Launch, WINFO, 2007 WiN-Canada conference 4 - Summary: - 14.6% of Canada's electricity is provided by Candu nuclear reactors; Nuclear equipment: 10 Research or isotope producing reactors - Pool-Type; Slowpoke 2; Sub-Critical assembly; NRU; and Maple; 22 Candu reactors providing electricity production - 18 of which are currently operating. Public acceptance: 41% feel nuclear should play more of a role, 67% support refurbishment, 48% support new build, 13% point gender gap in support, with men supporting more than women. Energy policy: Future of nuclear power - recognition that nuclear is part of the solution across Canada; New Build - 3 applications to regulator to prepare a site for new build, in Provinces of Ontario and Alberta, with one feasibility study underway in New Brunswick; Refurbishment - Provinces of Ontario (2010) and New Brunswick (2009). Nuclear waste management policy: Proposal submitted to regulator to prepare, construct and operate a deep geologic disposal facility in Ontario

Steps to Advanced CANDU 600

International Nuclear Information System (INIS)

Oh, Yongshick; Brooks, G. L.

1988-01-01

The CANDU nuclear power system was developed from merging of AECL heavy water reactor technology with Ontario Hydro electrical power station expertise. The original four units of Ontario Hydro's Pickering Generating Station are the first full-scale commercial application of the CANDU system. AECL and Ontario Hydro then moved to the next evolutionary step, a more advanced larger scale design for four units at the Bruce Generating Station. CANDU 600 followed as a single unit nuclear electric power station design derived from an amalgam of features of the multiple unit Pickering and Bruce designs. The design of the CANDU 600 nuclear steam supply system is based on the Pickering design with improvements derived from the Bruce design. For example, most CANDU 600 auxiliary systems are based on Bruce systems, whereas the fuel handling system is based on the Pickering system. Four CANDU 600 units are in operation, and five are under construction in Romania. For the additional four units at Pickering Generating Station 'B', Ontario Hydro selected a replica of the Pickering 'A' design with limited design changes to maintain a high level of standardization across all eight units. Ontario Hydro applied a similar policy for the additional four units at Bruce Generating Station 'B'. For the four unit Darlington station, Ontario Hydro selected a design based on Bruce with improvements derived from operating experience, the CANDU 600 design and development programs

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

International Nuclear Information System (INIS)

Scari, Maria Elizabeth

2017-01-01

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

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

Science.gov (United States)

Han, Jong-Hyun; Yoon, Do-Young

2010-06-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-07-15

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

CANDU-PHW fuel channel replacement experience

International Nuclear Information System (INIS)

Dunn, J.T.; Kakaria, B.K.

1982-09-01

One of the main characteristics of the CANDU pressurized heavy water reactor is the use of pressure tubes rather than one large pressure vessel to contain the fuel and coolant. This provides an inherent design capability to permit their replacement in an expeditious manner, without seriously affecting the high capacity factors of the reactor units. Of th eight Ontario Hydro commercial nuclear generating units, the lifetime performance places seven of them (including two that have had some of their fuel channels replaced), in the top ten positions in the world's large nuclear-electric unit performance ranking. Pressure tube cracks in the rolled joint region have resulted in 70 fuel channels being replaced in three reactor units, the latest being at the Bruce Nuclear Generating Station 'A', Unit 2 in February 1982. The rolled joint design and rolling procedures have been modified to eliminate this problem on CANDU units subsequent to Bruce 'A'. This paper describes the CANDU pressure tube performance history and expectations, and the tooling and procedures used to carry out the fuel channel replacement

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

KAUST Repository

Karsenty, Florent

2012-08-16

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Volume reduction/solidification of liquid radioactive waste using bitumen at Ontario Hydro's Bruce Nuclear Generating Station 'A'

International Nuclear Information System (INIS)

Day, J.E.; Baker, R.L.

1995-01-01

Ontario Hydro at the Bruce Nuclear Generating Station 'A' has undertaken a program to render the station's liquid radioactive waste suitable for discharge to Lake Huron by removing sufficient radiological and chemical contaminants to satisfy regulatory requirements for emissions. The system will remove radionuclide and chemical contaminants from five different plant waste streams. The contaminants will be immobilized and stored at on-site radioactive waste storage facilities and the purified streams will be discharged. The discharge targets established by Ontario Hydro are set well below the limits established by the Ontario Ministry of Environment (MOE) and are based on the Best Available Technology Economically Achievable Approach (B.A.T.E.A.). ADTECHS Corporation has been selected by Ontario Hydro to provide volume reduction/solidification technology for one of the five waste streams. The system will dry and immobilize the contaminants from a liquid waste stream in emulsified asphalt using thin film evaporation technology

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

International Nuclear Information System (INIS)

Chiba, Gou

2002-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2004-07-01

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

Top of tubesheet cracking in Bruce A NGS steam generator tubing - recent experience

International Nuclear Information System (INIS)

Clark, M.A.; Lepik, O.; Mirzai, M.; Thompson, I.

1998-01-01

During the Bruce A Nuclear Generating Station (BNGS-A) Unit 1 1997 planned outage, a dew point search method identified a leak in one steam generator(SG) tube. Subsequently, the tube was inspected with all available eddy current probes and removed for examination. The initial inspection results and metallurgical examination of the removed tube confirmed that the leak was due to intergranular attack/stress corrosion cracking (IGA/SCC) emanating from the secondary side of the tube at the top of the tubesheet location. Subsequently, eddy current and ultrasonic indications were found at the top of the tubesheet of other Alloy 600 SG tubes. To investigate the source of the indications and to validate the inspection probes, sections of 40 tubes with various levels of damage were removed. The metallurgical examination of the removed sections showed that both secondary side and primary side initiated, circumferential, stress corrosion cracking and intergranular attack occurred in the BNGS-A SG tubing. Significant degradation from both mechanisms was found, invariably located in the roll transition region of the top expansion joint between the tube and the tubesheet on the hot leg (304 degrees C) side of the tube. Various aspects of the failures and tube examinations are presented in this paper, including presentation of the cracking morphology, measured crack size distributions, and discussion of some factors possibly affecting the cracking. (author)

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

International Nuclear Information System (INIS)

2015-01-01

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

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

International Nuclear Information System (INIS)

Fiorina, Carlo; Hursin, Mathieu; Pautz, Andreas

2017-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-07-01

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

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

International Nuclear Information System (INIS)

Chang, Y.I.

1992-01-01

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

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

International Nuclear Information System (INIS)

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

1991-11-01

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

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

International Nuclear Information System (INIS)

Loika, E.F.

1994-01-01

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

Reactor Physics Programme

Energy Technology Data Exchange (ETDEWEB)

De Raedt, C

2000-07-01

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

Reactor Physics Programme

International Nuclear Information System (INIS)

De Raedt, C.

2000-01-01

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

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

Directory of Open Access Journals (Sweden)

Guangcan Zhu

2015-01-01

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

Nuclear reactor (1960)

International Nuclear Information System (INIS)

Maillard, M.L.

1960-01-01

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

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

International Nuclear Information System (INIS)

Behar, Christophe

2014-01-01

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

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

International Nuclear Information System (INIS)

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

2013-10-01

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

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

International Nuclear Information System (INIS)

Liu Wenjin; Zeng Zhongxiu; Ye Xianhui; Wu Wanjun

2013-01-01

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

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

International Nuclear Information System (INIS)

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

1996-09-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-09-01

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

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

International Nuclear Information System (INIS)

González Mantecón, Javier

2015-01-01

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

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.

Preliminary post-closure safety assessment of repository concepts for low level radioactive waste at the Bruce Site, Ontario

International Nuclear Information System (INIS)

Little, R.H.; Penfold, J.S.S.; Egan, M.J.; Leung, H.

2005-01-01

The preliminary post-closure safety assessment of permanent repository concepts for low-level radioactive waste (LLW) at the Ontario Power Generation (OPG) Bruce Site is described. The study considered the disposal of both short and long-lived LLW. Four geotechnically feasible repository concepts were considered (two near-surface and two deep repositories). An approach consistent with best international practice was used to provide a reasoned and comprehensive analysis of post-closure impacts of the repository concepts. The results demonstrated that the deep repository concepts in shale and in limestone, and the surface repository concept on sand should meet radiological protection criteria. For the surface repository concept on glacial till, it appears that increased engineering such as grouting of waste and voids should be considered to meet the relevant dose constraint. Should the project to develop a permanent repository for LLW proceed, it is expected that this preliminary safety assessment would need to be updated to take account of future site-specific investigations and design updates. (author)

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2009

International Nuclear Information System (INIS)

2014-02-01

JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. In the fiscal year 2009, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 6 cycles (daily operation : 24 days). The volume contains 138 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, prompt gamma-ray analyses, neutron activation analyses, RI productions, and others submitted by the users in JAEA and from other organizations. (author)

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2005

International Nuclear Information System (INIS)

2007-03-01

In the fiscal year 2005, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 37 cycles (daily operation : 137 days). JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography. Irradiation for activation analyses, radioisotope (RI) productions, fission tracks. Irradiation test of reactor materials etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT). Prompt gamma-ray analyses. Sensitivity measurement of radiation detectors. Experiment in the nuclear reactor training. Practice of Reactor operation. Irradiation for activation analyses, RI productions, fission tracks etc. The volume contains 100 activity reports, which are categorized into the fields of neutron scattering (9 subcategories), neutron radiography, neutron activation analyses, RI productions, prompt gamma-ray analyses, and others submitted by the users in JAEA and from other organizations. (author)

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2006

International Nuclear Information System (INIS)

2009-01-01

In the fiscal year 2006, the research reactor JRR-3 was operated 7 cycles (cycle operation: 26 days/cycle) for utilization sharing of the facility. And JRR-4 was operated 37 cycles (daily operation: 151 days). JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. The volume contains 294 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, neutron activation analyses, RI productions, prompt gamma-ray analyses, and others submitted by the users in JAEA and from other organizations. (author)

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2009

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-02-15

JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. In the fiscal year 2009, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 6 cycles (daily operation : 24 days). The volume contains 138 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, prompt gamma-ray analyses, neutron activation analyses, RI productions, and others submitted by the users in JAEA and from other organizations. (author)

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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

International Nuclear Information System (INIS)

Moukhamadeev, R.; Suvorov, A.

2000-01-01

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

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

International Nuclear Information System (INIS)

Okamoto, Koji

2015-01-01

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

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

Directory of Open Access Journals (Sweden)

New Pei Yee

2008-04-01

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

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

International Nuclear Information System (INIS)

Guidez, J.; Saturnin, A.

2016-01-01

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

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

International Nuclear Information System (INIS)

Zecevic, V.

1963-12-01

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

Safeguarding research reactors

International Nuclear Information System (INIS)

Powers, J.A.

1983-03-01

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

Reactor noise analysis of experimental fast reactor 'JOYO'

International Nuclear Information System (INIS)

Ohtani, Hideji; Yamamoto, Hisashi

1980-01-01

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

Volume reduction/solidification of liquid radioactive waste using bitumen at Ontario hydro's Bruce nuclear generating station open-quotes Aclose quotes

International Nuclear Information System (INIS)

Day, J.E.; Baker, R.L.

1994-01-01

Ontario Hydro at the Bruce Nuclear Generating Station open-quotes Aclose quotes has undertaken a program to render the station's liquid radioactive waste suitable for discharge to Lake Huron by removing sufficient radiological and chemical contaminants from five different plant waste streams. The contaminants will be immobilized and stored at on-site radioactive waste storage facilities and the purified streams will be discharged. The discharge targets established by Ontario Hydro are set well below the limits established by the Ontario Ministry of Environment (MOE) and are based on the Best Available Technology Economically Achievable Approach (B.A.T.E.A.). ADTECHS Corporation has been selected by Ontario Hydro to provide volume reduction/solidification technology for one of the five waste streams. The system will dry and immobilize the contaminants from a liquid waste stream in emulsified asphalt using thin film evaporation technology

Reactor safety study applied to the Forsmark 3 Power Plant

International Nuclear Information System (INIS)

Ericsson, G.; Tiren, L.I.

1978-01-01

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

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

International Nuclear Information System (INIS)

Kuscu, Ozlem Selcuk; Sponza, Delia Teresa

2009-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Nonboel, E

1985-07-01

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

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

International Nuclear Information System (INIS)

Matejka, Karel; Sklenka, Lubomir

1997-01-01

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

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)

TRIGA reactor main systems

International Nuclear Information System (INIS)

Boeck, H.; Villa, M.

2007-01-01

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

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

International Nuclear Information System (INIS)

Zhang Fan; Cai Zhangsheng; Yu Lei; Gui Xuewen

2005-01-01

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

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

International Nuclear Information System (INIS)

Takano, Hideki; Kaneko, Kunio.

1989-10-01

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

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2008

International Nuclear Information System (INIS)

2014-02-01

JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment and practice in the nuclear reactor training, Irradiation for activation analyses, RI productions, fission tracks etc. In the fiscal year 2008, the research reactor JRR-3 was operated for 7 cycles (cycle operation : 26days/cycle) for utilization sharing of facility. The research reactor JRR-4 was not operated in 2008. Because a crack was found on the weld of the aluminum cladding of a graphite reflector element. JRR-4 has remained shutdown until the reflector elements were replaced. The volume contains 250activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, neutron activation analyses, and others submitted by the users in JAEA and other Organizations. (author)

Generation IV reactors: economics

International Nuclear Information System (INIS)

Dupraz, B.; Bertel, E.

2003-01-01

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

Status of French reactors

International Nuclear Information System (INIS)

Ballagny, A.

1997-01-01

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

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

International Nuclear Information System (INIS)

Tuunanen, J.

1993-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1985-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

C. B. Davis

2006-07-01

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

Compact stellarators as reactors

International Nuclear Information System (INIS)

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

2001-01-01

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

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

International Nuclear Information System (INIS)

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

1993-01-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

UCLA program in reactor studies: The ARIES tokamak reactor study

International Nuclear Information System (INIS)

1991-01-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

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

International Nuclear Information System (INIS)

Mittag, S.

1994-06-01

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

Status of French reactors

Energy Technology Data Exchange (ETDEWEB)

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

1997-08-01

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

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2007

International Nuclear Information System (INIS)

2012-03-01

In the fiscal year 2007, the research reactor JRR-3 was operated for 7 cycles (cycle operation : 26days/cycle) and the JRR-4 was operated for 92 days. JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment and practice in the nuclear reactor training, Irradiation for activation analyses, RI productions, fission tracks, etc. The volume contains 262 activity reports, which are categorized into the fields of neutron scattering (10 subcategories), neutron radiography, neutron activation analyses, prompt gamma-ray analyses, and others submitted by the users in JAEA and other Organizations. (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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

International Nuclear Information System (INIS)

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

1983-05-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1966-07-01

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

Atomization of U3Si2 for research reactor fuel

International Nuclear Information System (INIS)

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

1995-01-01

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

Reactivity Coefficient Calculation for AP1000 Reactor Using the NODAL3 Code

Science.gov (United States)

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

2018-02-01

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

Advances in reactor physics education: Visualization of reactor parameters

International Nuclear Information System (INIS)

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

2012-01-01

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

DRAGON 3.05D, Reactor Cell Calculation System with Burnup

International Nuclear Information System (INIS)

2007-01-01

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

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

International Nuclear Information System (INIS)

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

1998-03-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Nuclear data usage for research reactors

International Nuclear Information System (INIS)

Nakano, Yoshihiro; Soyama, Kazuhiko; Amano, Toshio

1996-01-01

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

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

International Nuclear Information System (INIS)

Talamo, Alberto; Gudowski, Waclaw

2005-01-01

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

3 MW TRIGA Research Reactor facility of BAEC and its Utilization

International Nuclear Information System (INIS)

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

1999-01-01

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

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

International Nuclear Information System (INIS)

Jasiulevicius, A.

2001-01-01

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

Advanced reactor development: The LMR integral fast reactor program at Argonne

International Nuclear Information System (INIS)

Till, C.E.

1990-01-01

Reactor technology for the 21st Century must develop with characteristics that can now be seen to be important for the future, quite different from the things when the fundamental materials and design choices for present reactors were made in the 1950s. Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 3 figs

Development of Bundle Position-Wise Linear Model for Predicting the Pressure Tube Diametral Creep in CANDU Reactors

International Nuclear Information System (INIS)

Lee, Jae Yong; Na, Man Gyun

2011-01-01

Diametral creep of the pressure tube (PT) is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of a heat transport system. PT diametral creep leads to diametral expansion that affects the thermal hydraulic characteristics of the coolant channels and the critical heat flux. Therefore, it is essential to predict the PT diametral creep in CANDU reactors, which is caused mainly by fast neutron irradiation, reactor coolant temperature and so forth. The currently used PT diametral creep prediction model considers the complex interactions between the effects of temperature and fast neutron flux on the deformation of PT zirconium alloys. The model assumes that long-term steady-state deformation consists of separable, additive components from thermal creep, irradiation creep and irradiation growth. This is a mechanistic model based on measured data. However, this model has high prediction uncertainty. Recently, a statistical error modeling method was developed using plant inspection data from the Bruce B CANDU reactor. The aim of this study was to develop a bundle position-wise linear model (BPLM) to predict PT diametral creep employing previously measured PT diameters and HTS operating conditions. There are twelve bundles in a fuel channel and for each bundle, a linear model was developed by using the dependent variables, such as the fast neutron fluxes and the bundle temperatures. The training data set was selected using the subtractive clustering method. The data of 39 channels that consist of 80 percent of a total of 49 measured channels from Units 2, 3 and 4 were used to develop the BPLM models. The remaining 10 channels' data were used to test the developed BPLM models. The BPLM was optimized by the maximum likelihood estimation method. The developed BPLM to predict PT diametral creep was verified using the operating data gathered from the Units 2,3 and 4 in Korea. Two error components for the BPLM, which are the epistemic

Development of the environmental qualification safety requirement matrix for the containment system of in-service CANDU reactors

International Nuclear Information System (INIS)

Chun, R.M.; Low, J.; Sobolewski, J.

1994-01-01

Over the last several years, Ontario Hydro Nuclear (OHN) has placed increasing emphasis on environmental qualification (EQ) at its Pickering and Bruce NGS A and B nuclear generating stations (NGSs). The program currently underway (at the time of the conference) builds upon the experience gained from the extensive Darlington NGS EQ experience and from EQ programs conducted by other utilities. Some of the major steps of the OHN EQ program include: defining Safety Requirement Matrices (SRMs), establishing environmental conditions, developing an EQ List, conducting an EQ Assessment and maintaining Operational EQ Assurance during the plant life. The SRM identifies safety related components, their required safety functions and their mission times for each postulated design basis accident (DBA). This is a critical step, as the SRM defines the equipment that requires assurance of EQ and precise requirements must be provided to ensure a cost effective EQ program. This paper describes the development of the SRMs for the containment system of the Bruce stations. The introductory section briefly discusses how the industry has dealt with equipment qualification as it has evolved and the role of the SRMs in the OHN EQ Program. In Section 2, the preparation of the SRM is described along with the applicable ground rules used. The results of the application of the SRM preparation guidelines to the containment system are discussed in Section 3. A summary of the major findings and conclusions is presented. 3 refs., 3 figs

Analysis of a possible experimental assessment of a prototype fuel element containing burnable poison in the RA-3 reactor

International Nuclear Information System (INIS)

Lerner, Ana Maria; Madariaga, Marcelo

2002-01-01

The Argentine RA-3 research reactor (5 MW) is presently operated with LEU fuel by the National Atomic Energy Commission (CNEA). It belongs to the group of nuclear installations controlled, from the radiological and nuclear safety point of view, by the Nuclear Regulatory Authority (ARN). A new type of fuel elements containing burnable absorbers, with similar enrichment as the standard fuel elements but greater fissile contents, has recently been proposed for a new Argentine reactor design (RRR). In this framework the ARN considers interesting, if technically possible, the performance of an experiment in the RA-3 reactor. The experiment might enable, for such fuel element containing burnable poison, the verification of its neutronic behaviour under irradiation as well as a validation of the calculation line by comparison to measured values. It should be desirable that such experiment could reproduce as much as possible those conditions estimated for the RRR reactor, still under design in Argentina, having Silicide fuel elements with burnable poison, in the shape of cadmium wires in their structure. We here analyse a possible experiment consisting in the loading of a prototype fuel element with burnable poison in a normally loaded RA-3 core configuration. It would essentially be a standard RA-3 fuel element, having cadmium wires in its frame. This experiment would enable the verification of the prototype behaviour under irradiation, its operation limits and conditions, and particularly, the reactivity safety margins established in Argentine Standards, both calculated and measured. The main part of the experiment would imply some 200 full power days of operation at 5 MW, which would be drastically reduced if the reactor power is increased to 10 MW, as foreseen. We also show that under the proposed conditions, the experiment would not represent a significant penalty to the reactor normal operation. (author)

Reactor technology: power conversion systems and reactor operation and maintenance

International Nuclear Information System (INIS)

Powell, J.R.

1977-01-01

The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He 3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

Pilot plant production at Riso of LEU silicide fuel for the Danish reactor DR3

International Nuclear Information System (INIS)

Toft, P.; Borring, J.; Adolph, E.

1988-01-01

A pilot plant for fabricating LEU silicide fuel elements has been established at Riso National Laboratory. Three test elements for the Danish reactor DR3 have been fabricated, based on 19.88% enriched U 3 Si 2 powder that has been purchased elsewhere. The pilot plant has been set up and 3 test elements fabricated without any major difficulties

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

CANDU reactors with reactor grade plutonium/thorium carbide fuel

Energy Technology Data Exchange (ETDEWEB)

Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

2011-08-15

Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

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

Science.gov (United States)

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

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

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2007. Operation, utilization and technical development of JRR-3, JRR-4, NSRR and tandem accelerator

International Nuclear Information System (INIS)

Miyazaki, Osamu; Awa, Yasuaki; Isaka, Koji; Kutsukake, Kenichi; Komeda, Masao; Shibata, Ko; Hiyama, Kazuhisa; Suzuki, Mayu; Sone, Takuya; Ohuchi, Tomoaki; Terakado, Yuichi; Sataka, Masao

2009-06-01

The Department of Research Reactors and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor-3), JRR-4(Japan Research Reactor-4), NSRR(Nuclear Safety Research Reactor) and Tandem Accelerator. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2007 and March 31, 2008. The activities were categorized into five service/development fields: (1) Operation and maintenance of research reactors and tandem accelerator. (2) Utilization of research reactors and tandem accelerator. (3) Upgrading of utilization techniques of research reactors and tandem accelerator. (4) Safety administration for research reactors and tandem accelerator. (5) International cooperation. Also contained are lists of publications, meetings, granted permissions on lows and regulations concerning atomic energy, commendation, plans and outcomes in service and technical developments and so on. (author)

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2010. Operation, utilization and technical development of JRR-3, JRR-4, NSRR and Tandem Accelerator

International Nuclear Information System (INIS)

Ishii, Tetsuro; Nakamura, Kiyoshi; Kawamata, Satoshi; Yamada, Yusuke; Kawashima, Kazuhiro; Asozu, Takuhiro; Nakamura, Takemi; Arai, Masaji; Yoshinari, Shuji; Sataka, Masao

2012-03-01

The Department of Research Reactors and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor) and Tandem Accelerator. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2010 and March 31, 2011. The activities were categorized into five service/development fields: (1) Operation and maintenance of research reactors and tandem accelerator, (2) Utilization of research reactors and tandem accelerator, (3) Upgrading of utilization techniques of research reactors and tandem accelerator, (4) Safety administration for research reactors and tandem accelerator, (5) International cooperation. Also contained are lists of publications, meetings, granted permissions on lows and regulations concerning atomic energy, commendation, outcomes in service and technical developments and so on. (author)

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.)

Evaluation of the trial design studies for an advanced marine reactor, (3)

International Nuclear Information System (INIS)

Ambo, Noriaki; Yokomura, Takeyoshi.

1988-03-01

JAERI have carried out four core designs for three different type reactors (Semi-Integrated, Integrated and Integrated (self-pressured) type reactors), as the trial designs of an Advanced Marine Reactor for three years (1983 ∼ 1985). This report describes the result of comparison and studies of the core specific characteristics of these four cores, which include core concept, specifications, core life, specific power density, burn-up, reactivity control and etc. In conclusion, it was found that the Integrated type reactor core and the Semi-Integrated type reactor core designs satisfy the conditions of long core life (four years), high specific power density (50 ∼ 61 kw/l) and high burn-up (30,000 ∼ 32,000 MWD/t), so these two cores will be optimum designs based on the present technologies. (author)

Development of multi-physics code systems based on the reactor dynamics code DYN3D

Energy Technology Data Exchange (ETDEWEB)

Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany); Schuetze, Jochen [ANSYS Germany GmbH, Darmstadt (Germany); Frank, Thomas [ANSYS Germany GmbH, Otterfing (Germany); Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany)

2011-07-15

The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)

Development of multi-physics code systems based on the reactor dynamics code DYN3D

International Nuclear Information System (INIS)

Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich; Schuetze, Jochen; Frank, Thomas; Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo

2011-01-01

The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)

Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

International Nuclear Information System (INIS)

Jeong, Jae Jun; Chung, Bub Dong

2005-09-01

For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis

Impact of proposed research reactor standards on reactor operation

Energy Technology Data Exchange (ETDEWEB)

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

1974-07-01

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

Impact of proposed research reactor standards on reactor operation

International Nuclear Information System (INIS)

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

1974-01-01

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

Power measurement of the RA-3 reactor using the neutron noise technique and {sup 16}N; Medicion de la potencia del reactor RA-3, mediante la tecnica de ruido neutronico y nitrogeno 16

Energy Technology Data Exchange (ETDEWEB)

Gomez, Angel [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. Reactores y Centrales Nucleares

2003-07-01

This work describes a measurement method based on the neutron noise technique which is used for determining the relation between the power and the currents of two ionization chambers. These chambers are sensitive to the gamma radiation from the {sup 16}N decay produced in the RA-3 reactor core. The power during operation is obtained from the calibration factors by measuring those currents. As this calibration factors depend on the cooler flow that circulates in the reactor core and in the {sup 16}N measuring system, an estimator, that is a function of the ratio of this currents, is proposed in order to detect flow changes. (author)

INC '93 overview

International Nuclear Information System (INIS)

Howe, Bruce

1993-01-01

At the close of the International Nuclear Congress, INC '93, held in Toronto in October 1993, Bruce Howe, then president of AECL, presented a 'Review and Conclusions' under the following headings: Introduction, Future Energy Requirements, Economics of electricity generation, Radioactive waste management, Reactor safety, Public acceptance, Technology developments, Future energy strategies, Conclusion

MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

Energy Technology Data Exchange (ETDEWEB)

Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A. [AREVA - Tour AREVA, 1 Place Jean Millier, 92084 Paris La Defense (France)

2013-07-01

In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO{sub 2} fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory.

MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

International Nuclear Information System (INIS)

Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A.

2013-01-01

In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO 2 fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory

Fuel bundle impact velocities due to reverse flow

International Nuclear Information System (INIS)

Wahba, N.N.; Locke, K.E.

1996-01-01

If a break should occur in the inlet feeder or inlet header of a CANDU reactor, the rapid depressurization will cause the channel flow(s) to reverse. Depending on the gap between the upstream bundle and shield plug, the string of bundles will accelerate in the reverse direction and impact with the upstream shield plug. The reverse flow impact velocities have been calculated for various operating states for the Bruce NGS A reactors. The sensitivity to several analysis assumptions has been determined. (author)

Computational analysis of neutronic parameters for TRIGA Mark-II research reactor using evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3

International Nuclear Information System (INIS)

Altaf, M.H.; Badrun, N.H.; Chowdhury, M.T.

2015-01-01

Highlights: • SRAC-PIJ code and SRAC-CITATION have been utilized to model the core. • Most of the simulated results show no significant differences with references. • Thermal peak flux varies a bit due to up condition of TRIGA. • ENDF/B-VII.0 and JENDL-3.3 libraries perform well for neutronics analysis of TRIGA. - Abstract: Important kinetic parameters such as effective multiplication factor, k eff , excess reactivity, neutron flux and power distribution, and power peaking factors of TRIGA Mark II research reactor in Bangladesh have been calculated using the comprehensive neutronics calculation code system SRAC 2006 with the evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3. In the code system, PIJ code was employed to obtain cross section of the core cells, followed by the integral calculation of neutronic parameters of the reactor conducted by CITATION code. All the analyses were performed using the 7-group macroscopic cross section library. Results were compared to the experimental data, the safety analysis report (SAR) of the reactor provided by General Atomic as well as to the simulated values by numerically benchmarked MCNP4C, WIMS-CITATION and SRAC-CITATION codes. The maximum power densities at the hot spot were found to be 169.7 W/cc and 170.1 W/cc for data libraries ENDF/B-VII.0 and JENDL-3.3, respectively. Similarly, the total peaking factors based on ENDF/B-VII.0 and JENDL-3.3 were calculated as 5.68 and 5.70, respectively, which were compared to the original SAR value of 5.63, as well as to MCNP4C, WIMS-CITATION and SRAC-CITATION results. It was found in most cases that the calculated results demonstrate a good agreement with our experiments and published works. Therefore, this analysis benchmarks the code system and will be helpful to enhance further neutronics and thermal hydraulics study of the reactor

A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

International Nuclear Information System (INIS)

Guigon, B.; Vacelet, H.; Dornbusch, D.

2000-01-01

Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs from activation analysis to power reactor fuel qualification. In this paper the main characteristics of the Jules Horowitz Reactor are presented. Safety criteria are explained. Finally, merits and disadvantages of UMo compared to the standard U 3 Si 2 fuel are discussed. (author)

Multipurpose research reactors

International Nuclear Information System (INIS)

1988-01-01

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

Future fuel cycle and reactor strategies. Key issue paper no. 3

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-06-01

The scope of this paper includes those issues that are expected to remain or become important in the time period from 2015 to 2050. Events in this time frame are difficult to predict with any certainty, so the framework of this paper is necessarily somewhat speculative. The paper includes consideration of all facets of nuclear energy utilization, from ore extraction to final disposal of waste products. The paper first addresses the factors influencing the choice of reactor and fuel cycle. It then goes on to address the quantitatively largest category of reactor types expected to be important during the period; that is, thermal reactors burning uranium and plutonium fuel in various forms. The fast reactor type is then discussed both as stand-alone technology and as technology used in combination with thermal reactors. Thorium fuel use is discussed briefly. This paper is concentrated on the ``medium variant`` energy growth scenario identified in Key Issue Paper No. 1. The effects of either higher or lower growth could, of course, profoundly change the future development of the nuclear power industry. 31 refs, 5 figs, 4 tabs.

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2011. Operation, utilization and technical development of JRR-3, JRR-4, NSRR and tandem accelerator

International Nuclear Information System (INIS)

Ishii, Tetsuro; Nakamura, Kiyoshi; Kawamata, Satoshi; Ishikuro, Yasuhiro; Kawashima, Kazuhito; Kabumoto, Hiroshi; Nakamura, Takemi; Tamura, Itaru; Kawasaki, Sayuri; Sataka, Masao

2013-03-01

The Department of Research Reactors and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor) and Tandem Accelerator. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2011 and March 31, 2012. The activities were categorized into six service/development fields: (1) Recovery from the Great East Japan Earthquake, (2) Operation and maintenance of research reactors and tandem accelerator, (3) Utilization of research reactors and tandem accelerator, (4) Upgrading of utilization techniques of research reactors and tandem accelerator, (5) Safety administration for research reactors and tandem accelerator, (6) International cooperation. Also contained are lists of publications, meetings, granted permissions on lows and regulations concerning atomic energy, number of staff members dispatched to Fukushima for the technical assistance, commendation, outcomes in service and technical developments and so on. (author)

77 FR 65184 - Proposed Information Collection; Comment Request

Science.gov (United States)

2012-10-25

... sent to: Corporation for National and Community Service, Attn.: Bruce Kellogg, 3809C, 1201 New York... through Friday, except Federal holidays. (3) By fax to: (202) 606-3492, Attn.: Bruce Kellogg (4... FURTHER INFORMATION CONTACT: Bruce Kellogg, (202) 606-6954, or by email at [email protected

77 FR 65183 - Proposed Information Collection; Comment Request

Science.gov (United States)

2012-10-25

...: Corporation for National and Community Service, Attn.: Bruce Kellogg, 3809C, 1201 New York Avenue NW..., except Federal holidays. (3) By fax to: (202) 606-3492, Attn.: Bruce Kellogg. (4) Electronically through... INFORMATION CONTACT: Bruce Kellogg, (202) 606-6954, or by email at [email protected] . SUPPLEMENTARY...

76 FR 25673 - Proposed Information Collection; Comment Request

Science.gov (United States)

2011-05-05

...: Corporation for National and Community Service, National Service Trust; Attention Bruce Kellogg; 1201 New York... Friday, except Federal holidays. (3) By fax to: (202) 606-3492, Attention: Bruce Kellogg. (4..., Monday through Friday. FOR FURTHER INFORMATION CONTACT: Bruce Kellogg, (202) 606-6954, or by e-mail at...

76 FR 25672 - Proposed Information Collection; Comment Request

Science.gov (United States)

2011-05-05

... Community Service, National Service Trust; Attention Bruce Kellogg; 1201 New York Avenue, NW., Washington... holidays. (3) By fax to: (202) 606-3492, Attention: Bruce Kellogg. (4) Electronically through the... Friday. FOR FURTHER INFORMATION CONTACT: Bruce Kellogg, (202) 606-6954, or by e-mail at [email protected

Necessity of research reactors

International Nuclear Information System (INIS)

Ito, Tetsuo

2016-01-01

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

3-D thermal hydraulic analysis of transient heat removal from fast reactor core using immersion coolers

International Nuclear Information System (INIS)

Chvetsov, I.; Volkov, A.

2000-01-01

For advanced fast reactors (EFR, BN-600M, BN-1600, CEFR) the special complementary loop is envisaged in order to ensure the decay heat removal from the core in the case of LOF accidents. This complementary loop includes immersion coolers that are located in the hot reactor plenum. To analyze the transient process in the reactor when immersion coolers come into operation one needs to involve 3-D thermal hydraulics code. Furthermore sometimes the problem becomes more complicated due to necessity of simulation of the thermal hydraulics processes into the core interwrapper space. For example on BN-600M and CEFR reactors it is supposed to ensure the effective removal of decay heat from core subassemblies by specially arranged internal circulation circuit: 'inter-wrapper space'. For thermal hydraulics analysis of the transients in the core and in the whole reactor including hot plenum with immersion coolers and considering heat and mass exchange between the main sodium flow and sodium that moves in the inter-wrapper space the code GRIFIC (the version of GRIF code family) was developed in IPPE. GRIFIC code was tested on experimental data obtained on RAMONA rig under conditions simulating decay heat removal of a reactor with the use of immersion coolers. Comparison has been made of calculated and experimental result, such as integral characteristics (flow rate through the core and water temperature at the core inlet and outlet) and the local temperatures (at thermocouple location) as well. In order to show the capabilities of the code some results of the transient analysis of heat removal from the core of BN-600M - type reactor under loss-of-flow accident are presented. (author)

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

Comparison between TRU burning reactors and commercial fast reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

2001-03-01

additional consideration should be required in nuclear design and fuel treating facilities due to reactivity coefficient being shifted to the plus side, larger neutron yield and increased heat source caused by MA loading. (2) Confirmation of TRU burning reactor core concepts. The core specification of sodium cooled-nitride fueled TRU burning large reactor was designed based on commercial type fast reactor (sodium cooled nitride fueled large fast reactor, 38000 MWt) which was designed in the feasibility studies on commercialized fast reactor cycle system. The composition of MAs from LWR's spent fuel was supposed. MA content in the core fuel is settled to 60 wt% based on the JAERI's design in order to maximize the MA transmutation amount. We need to exchange 25% of core fuel with zirconium hydride (ZrH 1.6 ) to attain Doppler coefficient being equivalent to that of the conventional type commercial fast reactor loaded 5 wt% MA. Furthermore, this reactor could transmute MAs produced in forty-eight sodium cooled nitride fueled large fast reactors generating the same output. In order to investigate the dependency of MA transmutation characteristics on the reactor output, 1200 MWt TRU burning middle or small reactor core concept was designed. This core was settled by reducing the number of core fuel assemblies from that of TRU burning large reactor designed above. MA transmutation rate in this core is smaller than that in the TRU burning large reactor core because the neutron flux of this core becomes smaller than that of the TRU burning large reactor core due to the higher Pu enrichment. (3) Comparison between TRU burning reactor and conventional type commercial fast reactor. MA transmutation and nuclear characteristics of the sodium cooled nitride fuel commercial type fast reactor loaded 5 wt%MA were evaluated and compared with those of TRU burning large reactor designed in (2). The commercial type fast reactor could only transmute MAs produced in seven sodium cooled nitride

Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

International Nuclear Information System (INIS)

Joo, Young Sang; Lee, Jae Han

2007-02-01

The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification

Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

Energy Technology Data Exchange (ETDEWEB)

Joo, Young Sang; Lee, Jae Han

2007-02-15

The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification.

Calculation of fundamental parameters for the dynamical study of TRIGA-3-Salazar reactor (Mixed reactor core)

International Nuclear Information System (INIS)

Viais J, J.

1994-01-01

Kinetic parameters for dynamic study of two different configurations, 8 and 9, both with standard fuel, 20% enrichment and Flip (Fuel Life Improvement Program with 70% enrichment) fuel, for TRIGA Mark-III reactor from Mexico Nuclear Center, are obtained. A calculation method using both WIMS-D4 and DTF-IV and DAC1 was established, to decide which of those two configurations has the best safety and operational conditions. Validation of this methodology is done by calculate those parameters for a reactor core with new standard fuel. Configuration 9 is recommended to be use. (Author)