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.)

Criticality safety issues associated with the introduction of low void reactivity fuel in the Bruce reactors - a management and technical overview

International Nuclear Information System (INIS)

Thompson, J.W.; Austman, G.; Iglesias, F.; Schmeing, H.; Elliott, C.; Archinoff, G.

2004-01-01

The concept of criticality for operating reactor staff, particularly in a natural uranium-fuelled reactor, is relatively benign - the reactor is controlled at the critical condition by the regulating system. That is, issues related to criticality exist only within the reactor, in a set of carefully managed circumstances. With the introduction of enriched Low Void Reactivity Fuel (LVRF) into this operating environment comes a new 'concept of criticality', one which, although physically the same, cannot be treated in the same fashion. It may be the case that criticality can be achieved outside the reactor, albeit with a set of very pessimistic assumptions. Such 'inadvertent criticality' outside the reactor, should it occur, cannot be controlled. The consequences of such an inadvertent criticality could have far-reaching effects, not only in terms of severe health effects to those nearby, but also in terms of the negative impact on Bruce Power, and the Canadian nuclear industry in general. Thus the introduction of LVRF in the Bruce B reactors, and therefore the introduction of this new hazard, inadvertent criticality, warrants the development of a governance structure for its management. Such a program will consist of various elements, including the establishment of a framework to administer the criticality safety program, analytical assessment to support the process design, the development of operational procedures, the development of enhanced emergency procedures if necessary, and the implementation of a criticality safety training program. The entire package must be sufficient to demonstrate to station management, and the regulator, that the criticality safety risks associated with the implementation of enriched fuel have been properly evaluated, and that all necessary steps have been taken to effectively manage these risks. A well-founded Criticality Safety Program will offer such assurance. In this paper, we describe the establishment of a Criticality Safety

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

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

Clean energy for a new generation. Steam generator life cycle management and Bruce restart

International Nuclear Information System (INIS)

Newman, G.W.

2009-01-01

In the mid to late 1990s, Ontario Hydro decided to lay-up and write-down the Bruce A Nuclear Reactors. Upon transition to Bruce Power L.P., Canada's first and only private nuclear operator, new life and prospects were injected into the site, local economy and the provincial energy portfolio. The first step in this provincial power recovery initiative involved restart of Bruce Units 3 and 4 in the 2003/04 time-frame. Units 3 and 4 have performed beyond expectation during the last five-year operating interval. A combination of steam generator and fuel channel issues precluded a similar restart of Units 1 and 2. Enter the refurbishment of Bruce Units 1 and 2. This first-of-a-kind undertaking within the Canadian nuclear power industry is testament to the demonstrated industry leadership by Bruce Power L.P., their investors and the significant vendor community contribution that is supporting this major power infrastructure enhancement. Initiated as a 'turn-key' project solution separated from the operating units, this major refurbishment project has evolved to a fully managed in-house refurbishment project with the continued support from the broader vendor community. As part of this first-of-kind undertaking, Bruce Power L.P. is in the process of accomplishing such initiatives as a complete fuel channel re-tube (i.e. full core calandria and pressure tube replacement), replacement of all boilers (i.e. 16 in total) and the majority of feeder pipe replacement. Complimentary major upgrades and replacement of the remainder of plant equipment including both nuclear and non-nuclear valves, heat exchangers, electrical infrastructure, service water systems and components, all while meeting a parallel evolving/maturing regulatory environment related to achieving compliance with IAEA derived modern codes and standards. Returning to ground level, boiler replacement is a key part of the refurbishment undertaking and this further reflected a meeting of the 'old' and the 'new'. Pre

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

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

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)

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)

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.

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)

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)

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

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)

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)

Fuel string supporting shield plug (f3sp) for Ontario Hydro - Bruce NGSA

Energy Technology Data Exchange (ETDEWEB)

Henry, P T [Canadian General Electric Co. Ltd., Peterborough, ON (Canada)

1997-12-31

A reactor `power pulse` problem was identified for the Ontario Hydro Bruce generating stations. On a postulated inlet header break, the fuel strings in a large number of channels could relocate toward the upstream end, resulting in a power pulse. The solution adopted for Bruce GSA is to change the direction of fuelling, from against the flow, to fuelling with the flow. In this revised fuelling scheme, given a postulated inlet header failure, the fuel bundle with the highest burnup would relocate into the reactor core and introduce a negative reactivity during the accident. However, this fuelling configuration results in a highly irradiated fuel bundle residing in the most downstream position against the latch. The latch supports only the outer ring of elements, not the end plate. A resulting high stress on the end plate coupled with high levels of hydrogen and deuterium may result in Zr hydride assisted cracking in the end plate during hot shutdown conditions. (In fuelling against flow, this is not a problem, since the latch supported bundle is not irradiated and has only low levels of hydrogen and deuterium.) A fuel string supporting shield plug (f3sp) which supports the bundle end plate has been developed as a solution to the fuel bundle end plate cracking problem. It would replace the existing outlet shield plug in all channels. This paper will describe the f3sp design, associated fuel handling, operation and qualification for reactor use. (author). 8 figs.

Fuel string supporting shield plug (f3sp) for Ontario Hydro - Bruce NGSA

International Nuclear Information System (INIS)

Henry, P.T.

1996-01-01

A reactor 'power pulse' problem was identified for the Ontario Hydro Bruce generating stations. On a postulated inlet header break, the fuel strings in a large number of channels could relocate toward the upstream end, resulting in a power pulse. The solution adopted for Bruce GSA is to change the direction of fuelling, from against the flow, to fuelling with the flow. In this revised fuelling scheme, given a postulated inlet header failure, the fuel bundle with the highest burnup would relocate into the reactor core and introduce a negative reactivity during the accident. However, this fuelling configuration results in a highly irradiated fuel bundle residing in the most downstream position against the latch. The latch supports only the outer ring of elements, not the end plate. A resulting high stress on the end plate coupled with high levels of hydrogen and deuterium may result in Zr hydride assisted cracking in the end plate during hot shutdown conditions. (In fuelling against flow, this is not a problem, since the latch supported bundle is not irradiated and has only low levels of hydrogen and deuterium.) A fuel string supporting shield plug (f3sp) which supports the bundle end plate has been developed as a solution to the fuel bundle end plate cracking problem. It would replace the existing outlet shield plug in all channels. This paper will describe the f3sp design, associated fuel handling, operation and qualification for reactor use. (author). 8 figs

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)

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).

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)

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 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 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...

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)

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)

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.

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)

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].

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

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)

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

9 CFR 78.7 - Brucellosis reactor cattle.

Science.gov (United States)

2010-01-01

... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Brucellosis reactor cattle. 78.7... Restrictions on Interstate Movement of Cattle Because of Brucellosis § 78.7 Brucellosis reactor cattle. (a) Destination. Brucellosis reactor cattle may be moved interstate only for immediate slaughter as follows: (1...

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)

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

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.

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.

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)

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)

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)

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

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.

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

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)

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)

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

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

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)

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)

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

International Nuclear Information System (INIS)

Kalcheva, Silva; Koonen, Edgar

2011-01-01

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

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)

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)

Degradation of Acid Orange 7 Dye in Two Hybrid Plasma Discharge Reactors

Science.gov (United States)

Shen, Yongjun; Lei, Lecheng; Zhang, Xingwang; Ding, Jiandong

2014-11-01

To get an optimized pulsed electrical plasma discharge reactor and to increase the energy utilization efficiency in the removal of pollutants, two hybrid plasma discharge reactors were designed and optimized. The reactors were compared via the discharge characteristics, energy transfer efficiency, the yields of the active species and the energy utilization in dye wastewater degradation. The results showed that under the same AC input power, the characteristics of the discharge waveform of the point-to-plate reactor were better. Under the same AC input power, the two reactors both had almost the same peak voltage of 22 kV. The peak current of the point-to-plate reactor was 146 A, while that of the wire-to-cylinder reactor was only 48.8 A. The peak powers of the point-to-plate reactor and the wire-to-cylinder reactor were 1.38 MW and 1.01 MW, respectively. The energy per pulse of the point-to-plate reactor was 0.2221 J, which was about 29.4% higher than that of the wire-to-cylinder reactor (0.1716 J). To remove 50% Acid Orange 7 (AO7), the energy utilizations of the point-to-plate reactor and the wire-to-cylinder reactor were 1.02 × 10-9 mol/L and 0.61 × 10-9 mol/L, respectively. In the point-to-plate reactor, the concentration of hydrogen peroxide in pure water was 3.6 mmol/L after 40 min of discharge, which was higher than that of the wire-to-cylinder reactor (2.5 mmol/L). The concentration of liquid phase ozone in the point-to-plate reactor (5.7 × 10-2 mmol/L) was about 26.7% higher than that in the wire-to-cylinder reactor (4.5 × 10-2 mmol/L). The analysis results of the variance showed that the type of reactor and reaction time had significant impacts on the yields of the hydrogen peroxide and ozone. The main degradation intermediates of AO7 identified by gas chromatography and mass spectrometry (GCMS) were acetic acid, maleic anhydride, p-benzoquinone, phenol, benzoic acid, phthalic anhydride, coumarin and 2-naphthol. Proposed degradation pathways were

Degradation of Acid Orange 7 Dye in Two Hybrid Plasma Discharge Reactors

International Nuclear Information System (INIS)

Shen Yongjun; Ding Jiandong; Lei Lecheng; Zhang Xingwang

2014-01-01

To get an optimized pulsed electrical plasma discharge reactor and to increase the energy utilization efficiency in the removal of pollutants, two hybrid plasma discharge reactors were designed and optimized. The reactors were compared via the discharge characteristics, energy transfer efficiency, the yields of the active species and the energy utilization in dye wastewater degradation. The results showed that under the same AC input power, the characteristics of the discharge waveform of the point-to-plate reactor were better. Under the same AC input power, the two reactors both had almost the same peak voltage of 22 kV. The peak current of the point-to-plate reactor was 146 A, while that of the wire-to-cylinder reactor was only 48.8 A. The peak powers of the point-to-plate reactor and the wire-to-cylinder reactor were 1.38 MW and 1.01 MW, respectively. The energy per pulse of the point-to-plate reactor was 0.2221 J, which was about 29.4% higher than that of the wire-to-cylinder reactor (0.1716 J). To remove 50% Acid Orange 7 (AO7), the energy utilizations of the point-to-plate reactor and the wire-to-cylinder reactor were 1.02 × 10 −9 mol/L and 0.61 × 10 −9 mol/L, respectively. In the point-to-plate reactor, the concentration of hydrogen peroxide in pure water was 3.6 mmol/L after 40 min of discharge, which was higher than that of the wire-to-cylinder reactor (2.5 mmol/L). The concentration of liquid phase ozone in the point-to-plate reactor (5.7 × 10 −2 mmol/L) was about 26.7% higher than that in the wire-to-cylinder reactor (4.5 × 10 −2 mmol/L). The analysis results of the variance showed that the type of reactor and reaction time had significant impacts on the yields of the hydrogen peroxide and ozone. The main degradation intermediates of AO7 identified by gas chromatography and mass spectrometry (GCMS) were acetic acid, maleic anhydride, p-benzoquinone, phenol, benzoic acid, phthalic anhydride, coumarin and 2-naphthol. Proposed degradation

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

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

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)

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)

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)

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.

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

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

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

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.

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)

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…

Diagrams about RA reactor operation, Annex 7; Prilog 7 - Rad reaktora RA po godinama - dijagrami

Energy Technology Data Exchange (ETDEWEB)

Martinc, R [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1976-01-15

This Annex 7 includes diagrams about RA reactor operation (MWh) from 1960-1975; mean values of reactor power per day and month in 1975, and percent of utilisation of experimental space during 1975. [Serbo-Croat] Ovaj prilog sadrzi dijagrame o radu reaktora (MWh) po godinama (1960-1975); srednje vrednosti dnevne snage reaktora u 1975. godini; rad reakrora (MWh) po mesecima u 1975; procenat iskoriscenja eksperimentalnog prostora u 1975. godini.

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)

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 3, Sessions 12-16

Energy Technology Data Exchange (ETDEWEB)

Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)

1995-09-01

This document, Volume 3, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, ad the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected abstracts have been indexed separately for inclusion in the Energy Science and Technology Database.

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

International Nuclear Information System (INIS)

Block, R.C.; Feiner, F.

1995-09-01

This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 2, Sessions 6-11

Energy Technology Data Exchange (ETDEWEB)

Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)

1995-09-01

This document, Volume 2, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

Energy Technology Data Exchange (ETDEWEB)

Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)

1995-09-01

This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

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

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

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

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

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)

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

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.

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.

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)

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....

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.

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

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

Status of IVO-FR2-Vg7 experiment for irradiation of fast reactor fuel rods

International Nuclear Information System (INIS)

Elbel, H.; Kummerer, K.; Bojarsky, K.; Lopez Jimenez, J.; Otero de la Gandara, J.L.

1979-01-01

Report on the Seminar celebrated in Madrid between KfK (Karlsruhe) and JEN (Madrid) concerning a Joint Irradiation Program of Fast Reactor Fuel Rods. The design of fuel rods in general is defined, and, in particular of those with a density 94% DT and diameter 7.6 mm up to a burn-up of 7% FIMA, to be irradiated in the FR2 Reactor (Karlsruhe). Together with the design of NaK and single-wall capsules used in this irradiation, other possibilities of irradiation in the reactor will also be described. (auth.)

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

Enhanced decolourisation of Acid Orange 7 in a continuous UASB reactor with quinones as redox mediators.

NARCIS (Netherlands)

Cervantes, F.J.; Zee, van der F.P.; Lettinga, G.; Field, J.A.

2001-01-01

The reductive biotransformation of acid orange 7 (AO7) was explored in a lab-scale upflow anaerobic sludge blanket (UASB) reactor at low hydraulic residence times (HRT). A colour removal of 85% was achieved when the reactor was operated at a HRT of 6 hours, but decreased up to 70% when the HRT was

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

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

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)

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)

Summary report of the 7th reduced-moderation water reactor workshop

International Nuclear Information System (INIS)

Akie, Hiroshi; Nabeshima, Kunihiko; Uchikawa, Sadao

2005-08-01

As a research on the future innovative water reactor, the development of Reduced-Moderation Water Reactors (RMWRs) has been performed in Japan Atomic Energy Research Institute (JAERI). The workshop on RMWRs is aiming at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors, and has been held every year since 1998. The 7th workshop was held on March 5, 2004 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The program of the workshop was composed of 5 lectures and an overall discussion time. The workshop started with the lecture by JAERI on the status and future program of PMWR research and development, followed by the two presentations by JAERI and Japan Nuclear Cycle Development Institute, respectively, on the investigation and evaluation of water cooled reactor in Feasibility Study Program on Commercialized Fast Reactor Systems. The lectures were also made on the Japan's nuclear fuel cycle and scenarios for RMWRs deployment by JAERI, and on the next generation reactor development activity by Hitachi, Ltd. The main subjects of the overall discussion time were Na cooled fast reactor, deployment effects of RMWRs and the future plan of the RMWR research and development. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as of the discussion time. In addition in the Appendices, there are included presentation handouts of each lecture, program of the workshop and the participants list. (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)

Research reactors - an overview

International Nuclear Information System (INIS)

West, C.D.

1997-01-01

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

An approach of raising the low power reactor trip block (P-7) in Maanshan Power Plant

International Nuclear Information System (INIS)

Wang, L.C.

1984-01-01

The technical specification for the Maanshan Nuclear Power Station (FSAR Table 16.2.2-3) requires that with an increasing reactor power level above the setpoint of low power reactor trip block (P-7), a turbine trip shall initiate a reactor trip. This anticipatory reactor trip on turbine trip prevents the pressurizer PORV from openning during turbine trip event. In order to reduce unnecessary reactor trip due to turbine trip on low reactor power level during Maanshan start-up stage, Taiwan Power Company performed a transient analysis for turbine trip event by using RETRAN code. The highest reactor power level at which a turbine trip will not open the pressurizer PORV is searched. The results demonstrated that this power level can be increased from the original value-10% of the rated thermal power-to about 48% of the rated thermal power

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)

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.

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)

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)

Syngas fermentation by Clostridium carboxidivorans P7 in a horizontal rotating packed bed biofilm reactor with enhanced ethanol production

International Nuclear Information System (INIS)

Shen, Yanwen; Brown, Robert C.; Wen, Zhiyou

2017-01-01

Highlights: • A novel a horizontal rotating packed bed (h-RPB) reactor for syngas fermentation was reported. • The h-RPB reactor enhanced ethanol productivity by 3.3-folds compared to continuous stirred tank reactor (CSTR). • The h-RPB reactor has a unique feature of transfer gas from both bulk liquid phase and headspace phase. • The mass transfer in the headspace of h-PRB played an important role for enhanced ethanol production. - Abstract: Gasification of lignocellulosic biomass followed by syngas fermentation is a promising process for producing fuels and chemicals. Syngas fermentation, however, is commonly limited by low mass transfer rates. In this work, a horizontally oriented rotating packed bed (h-RPB) reactor was developed to improve mass transfer and enhance ethanol production. In the h-RPB reactor, cell attachment materials were packed in the reactor and half submerged in the liquid and half exposed to the headspace. With continuous rotation of the packing materials, the cells in biofilm were alternately in contact with liquid and headspace; thus, transport of syngas to the cells occurred in both the liquid phase and headspace. The volumetric mass transfer coefficient (k_La) of the h-RPB reactor was lower than that in a traditional continuous stirred tank reactor (CSTR), indicating the mass transfer in the liquid phase of h-PRB was lower than CSTR, and the mass transfer in the headspace phase played an important role in syngas fermentation. The syngas fermentation of Clostridium carboxidivorans P7 in h-RPB resulted in a 7.0 g/L titer and 6.7 g/L/day productivity of ethanol, respectively, 3.3 times higher than those obtained in a CSTR under the same operational conditions. The results demonstrate that the h-RPB reactor is an efficient system for syngas fermentation, making cellulosic ethanol biorefinery one step closer to technical and economic feasibility.

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.)

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

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.)

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

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.

IGORR 7: Proceedings of the 7. meeting of the International Group On Research Reactors

International Nuclear Information System (INIS)

1999-01-01

According to the subjects covered the papers presented at the meeting were divided into following sessions: New research reactor projects; secondary neutron sources; New research reactor facilities; Improvement of Research Reactors Facilities; Research and Development Needs

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

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.

Light water reactor fuel analysis code FEMAXI-7. Model and structure

International Nuclear Information System (INIS)

Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa; Saitou, Hiroaki

2013-07-01

A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in both normal conditions and anticipated transient conditions. This code is an advanced version which has been produced by incorporating the former version FEMAXI-6 with numerous functional improvements and extensions. In FEMAXI-7, many new models have been added and parameters have been clearly arranged. Also, to facilitate effective maintenance and accessibility of the code, modularization of subroutines and functions have been attained, and quality comment descriptions of variables or physical quantities have been incorporated in the source code. With these advancements, the FEMAXI-7 code has been upgraded to a versatile analytical tool for high burnup fuel behavior analyses. This report describes in detail the design, basic theory and structure, models and numerical method of FEMAXI-7, and its improvements and extensions. (author)

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

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

International Nuclear Information System (INIS)

1985-09-01

The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

Light water reactor fuel analysis code FEMAXI-7; model and structure

International Nuclear Information System (INIS)

Suzuki, Motoe; Udagawa, Yutaka; Saitou, Hiroaki

2011-03-01

A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in both normal conditions and anticipated transient conditions. This code is an advanced version which has been produced by incorporating the former version FEMAXI-6 with numerous functional improvements and extensions. In FEMAXI-7, many new models have been added and parameters have been clearly arranged. Also, to facilitate effective maintenance and accessibility of the code, modularization of subroutines and functions have been attained, and quality comment descriptions of variables or physical quantities have been incorporated in the source code. With these advancements, the FEMAXI-7 code has been upgraded to a versatile analytical tool for high burnup fuel behavior analyses. This report describes in detail the design, basic theory and structure, models and numerical method, and improvements and extensions. (author)

Operating performance and reliability of CANDU PHWR fuel channels in Canada

International Nuclear Information System (INIS)

Strachan, B.; Brown, D.R.

1983-03-01

CANDU nuclear plants use many small-diameter high-pressure fuel channels. Good operating performance from the CANDU fuel channels has made a major contribution to the world-leading operating record of the CANDU nuclear power plants. As of 1982 December 31, there were 7,480 fuel channels installed in 18 CANDU reactors over 500 MW(e) in size. Eight of these reactors have been declared in-service and have accumulated 24,000 fuel channel-years of operation. The only significant operating problems with fuel channels have been the occurrence of leaking cracks in 70 fuel channels and a larger amount of axial creep on the early reactors than was originally provided for in the design. Both of these problems have been corrected on all CANDU reactors built since the Bruce GS 'A' station and the newer reactors should exhibit even better performance

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

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

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.

Project management lessons learned from building the Wendelstein 7-x stellerator fusion research reactor

Energy Technology Data Exchange (ETDEWEB)

Freire-Gormaly, M.; Gittens, A.; Zhang, L., E-mail: m.freire.gormaly@utoronto.ca, E-mail: antonio.gittens@mail.utoronto.ca, E-mail: lavender.zhang@outlook.com [University of Toronto, Toronto, ON (Canada)

2015-07-01

Wendelstein 7-X (W7-X) is the world's largest 'stellerator' nuclear fusion reactor being commissioned in Greifswald, Germany. It will inform the international fusion energy test device (ITER). The complexity of W7- X added challenges since industrial expertise to manufacture components did not exist. The construction was completed eight years behind schedule and almost 100% over budget. Key take-away lessons in project management were revealed from W7-X which can be applied to any nuclear project. These lessons are aligned with the project management knowledge areas of schedule, stakeholder, procurement, scope, schedule, cost, communication, risk, quality, human resources and procurement management. (author)

Project management lessons learned from building the Wendelstein 7-x stellerator fusion research reactor

International Nuclear Information System (INIS)

Freire-Gormaly, M.; Gittens, A.; Zhang, L.

2015-01-01

Wendelstein 7-X (W7-X) is the world's largest 'stellerator' nuclear fusion reactor being commissioned in Greifswald, Germany. It will inform the international fusion energy test device (ITER). The complexity of W7- X added challenges since industrial expertise to manufacture components did not exist. The construction was completed eight years behind schedule and almost 100% over budget. Key take-away lessons in project management were revealed from W7-X which can be applied to any nuclear project. These lessons are aligned with the project management knowledge areas of schedule, stakeholder, procurement, scope, schedule, cost, communication, risk, quality, human resources and procurement management. (author)

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.”

Design and construction of a 7,500 liter immobilized cell reactor-separator for ethanol production from whey

Energy Technology Data Exchange (ETDEWEB)

Dale, M.C.

1992-12-31

A 7,500 liter reactor/separator has been constructed for the production of ethanol from concentrated whey permeate. This unit is sited in Hopkinton IA, across the street from a whey generating cheese plant A two phase construction project consisting of (1) building and testing a reactor/separator with a solvent absorber in a single unified housing, and (2) building and testing an extractive distillation/product stripper for the recovery of anhydrous ethanol is under way. The design capacity of this unit is 250,000 gal/yr of anhydrous product. Design and construction details of the reactor/absorber separator are given, and design parameters for the extractive distillation system are described.

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

Digital reactor period meter type of NSSG-7

Energy Technology Data Exchange (ETDEWEB)

Glowacki, S W

1981-01-01

The paper presents the idea and electronic circuits of the Digital Reactor Period Meter. The instrument consists of a neutron ionisation chamber, the amplifier logarithming the output chamber current, the circuit taking two samples of the log amplifier output signal and subtracting them, the analog -to -digital dividing circuit and the scaler providing the final information of the reactor period value in seconds and in the digital form. Besides it, the instrument produces the acoustic signal in the case, when the rise-time of neutron flux exceeds the permitted value. The untypical construction of the reactor period meter has been developed to obtain both good measurement accuracy and the resistance against the electromagnetic background pulses interfering with the measuring process. The applied measuring system has been patented.

PLUS7TM In-Reactor Operating Performance and Economics

International Nuclear Information System (INIS)

Kim, Kyutae; Jang, Youngki; Choi, Joonhyung; Lee, Jinseok; Kim, Yoonho; Suh, Jungmin

2006-01-01

KNFC has developed an advanced fuel, PLUS7 TM , for the Korean Standard Nuclear Power Plants(KSNPs) through the joint development program with Westinghouse. With the help of various out-of-pile tests, it is found that the PLUS7 TM shows much better performance than the current fuel, GUARDIAN TM from the safety and economy points of view. Now four Lead Test Assembles(LTAs) of the PLUS7 TM are being irradiated for the 3 rd cycle after the successful completion of the 1 st and 2 nd irradiation cycles. During the 1 st and 2 nd irradiation cycles, no fuel failure was observed at LTAs and their nuclear-related parameters matched their design values well. During the overhaul period, on the other hand, pool side examinations were performed for four LTAs to generate key in-reactor fuel performance data such as fuel rod and assembly growths, fuel rod-to-top nozzle gap, fuel assembly bow and twist, fuel rod bow, spacer grid width, fuel rod diameter and fuel rod oxide layer thickness. It is found that all measured values are bounded by upper and lower predicted ones. The detailed economic analyses have shown that significant fuel cycle cost can be reduced by more than one million dollars per cycle of one KSNP with the introduction of the PLUS7 TM assembly. Furthermore, more than one hundred million dollars with power up-rating of 5% can be saved annually for currently operating eight KSNPs, which is easily and safety achievable with the PLUS7 TM assembly

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)

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

7. European conference of TRIGA reactor users. Conference papers

Energy Technology Data Exchange (ETDEWEB)

NONE

1982-07-01

At the Seventh European Conference of TRIGA Users, held in September 1982, in Istanbul, Turkey, the following aspects are discussed: safety aspects of TRIGA reactors; developments and improvements; operating and maintenance experiences; applications; reactor calculations; fuel cycle aspects and research programs.

7. European conference of TRIGA reactor users. Conference papers

International Nuclear Information System (INIS)

1982-01-01

At the Seventh European Conference of TRIGA Users, held in September 1982, in Istanbul, Turkey, the following aspects are discussed: safety aspects of TRIGA reactors; developments and improvements; operating and maintenance experiences; applications; reactor calculations; fuel cycle aspects and research programs

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.

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

Reconstruction of intra-bundle fission density profile during a postulated LOCA in a CANDU reactor

Energy Technology Data Exchange (ETDEWEB)

Ilas, D. [Oak Ridge National Laboratory (United States); Rahnema, F. [Georgia Inst. of Technology (United States); Nuclear and Radiological Engineering/Medical Physics Programs, George W. Woodruff School, Georgia Inst. of Technology, Atlanta, GA 30332-0405 (United States); Serghiuta, D. [Canadian Nuclear Safety Commission (Canada); Sarsour, H.; Turinsky, P. J. [North Carolina State Univ. (United States); Stamm' ler, R. [Studsvik Scandpower AS (Norway)

2006-07-01

In this paper, results related to the reconstruction of intra-bundle fission density profile for a 37-pin CANDU-6 bundle with the highest enthalpy deposition during a postulated large LOCA stagnation break in a Bruce B core are presented. Bruce B is a nuclear power plant in Kincardine, Ontario (Canada)), on the shores of Lake Huron with 4 CANDU reactors that are rated at about 750 MWe. The reconstruction of the fuel pin fission densities is based on steady-state, three-dimensional simulations with the Monte Carlo code MCNP for a subset of 27 out of 69 time steps during the first two seconds of the power pulse predicted for the fuel bundle at core location V13/8. Two-group cross section data libraries are generated for MCNP at each time step by the lattice depletion neutron transport code HELIOS-1.7. To include the effect of the surrounding core environment, the calculations are performed with time-dependent albedo boundary conditions inferred from a full core simulation of the transient by the nodal diffusion code NESTLE with HELIOS homogenized cross-sections. It is found that the local peaking factor (LPF) in the outer ring varies during the transient, but never exceeds its value before the transient. Inclusion of the core environment increases the LPF in the outer ring. For the analyzed case, the increase is 0.72% with a relative error of 0.01% for the LPF before the transient and 0.55% (with a relative error of 0.01%) for the maximum average LPF during the transient. The latter is based on only four selected transient time points. Note that the immediate environment of the 'hot bundle' does not contain any reactivity devices or other perturbing factors. As a result, the increases observed in the LPF in the outer ring may not be representative of the situations in which 'other' core environment perturbing factors are present. To determine the effect of these factors on the LPF, further analyses of a bundle in the proximity of control devices

Advances in Reactor Physics, Mathematics and Computation. Volume 2

Energy Technology Data Exchange (ETDEWEB)

1987-01-01

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

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.

Reactor of the XXI century

International Nuclear Information System (INIS)

Zhotabaev, Zh.R.; Solov'ev, Yu.A.

2001-01-01

The advantages of both molten salt reactors (MSR) and homogenous molten salt reactors (HMSR) are illuminated. It is noted that the MSR possess accident probability A=10 -6 1/reactor.years and the HMSR with integral configuration has A=10 -7 1/reactor.years. The methods for these reactors technological problems solution - tritium removal, salt melt circulation and capacity pick up - are discussed

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)

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

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Sessions 17-24

Energy Technology Data Exchange (ETDEWEB)

Block, R.C.; Feiner, F. [American Nuclear Society, La Grange Park, IL (United States)

1995-09-01

Technical papers accepted for presentation at the Seventh International Topical Meeting on Nuclear Reactor Thermal-Hydraulics are included in the present Proceedings. Except for the invited papers in the plenary session, all other papers are contributed papers. The topics of the meeting encompass all major areas of nuclear thermal-hydraulics, including analytical and experimental works on the fundamental mechanisms of fluid flow and heat transfer, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Because of the complex nature of nuclear reactors and power plants, several papers deal with the combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. The participation in the conference by the authors from several countries and four continents makes the Proceedings a comprehensive review of the recent progress in the field of nuclear reactor thermal-hydraulics worldwide. Individual papers have been cataloged separately.

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Sessions 17-24

International Nuclear Information System (INIS)

Block, R.C.; Feiner, F.

1995-09-01

Technical papers accepted for presentation at the Seventh International Topical Meeting on Nuclear Reactor Thermal-Hydraulics are included in the present Proceedings. Except for the invited papers in the plenary session, all other papers are contributed papers. The topics of the meeting encompass all major areas of nuclear thermal-hydraulics, including analytical and experimental works on the fundamental mechanisms of fluid flow and heat transfer, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Because of the complex nature of nuclear reactors and power plants, several papers deal with the combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. The participation in the conference by the authors from several countries and four continents makes the Proceedings a comprehensive review of the recent progress in the field of nuclear reactor thermal-hydraulics worldwide. Individual papers have been cataloged separately

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

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

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)

Monte Carlo simulation of a research reactor with nominal power of 7 MW to design new control safety rods

Energy Technology Data Exchange (ETDEWEB)

Shoushtari, M.K.; Kakavand, T. [Faculty of Science, University of Zanjan, Zanjan, P.O. Box 451-313 (Iran, Islamic Republic of); Sadat Kiai, S.M., E-mail: sadatkiai@yahoo.co [Nuclear Science and Technology Research Institute (NSTR), Nuclear Science Research, A.E.O.I., P.O. Box 14155-1339, Tehran (Iran, Islamic Republic of); Ghaforian, H. [Faculty of Science and Technology of Marine, Tehran (Iran, Islamic Republic of)

2010-03-01

The Monte Carlo simulation has been established for a research reactor with nominal power of 7 MW. A detailed model of the reactor core was employed including standard and control fuel elements, reflectors, irradiation channels, control rods, reactor pool and thermal column. The following physical parameters of reactor core were calculated for the present LEU core: core reactivity (rho), control rod (CR) worth, thermal and epithermal neutron flux distributions, shutdown margin and delayed neutron fraction. Reduction of unfavorable effects of blockage probability of control safety rod (CSR)s in their interiors because of not enough space in their sites, and lack of suitable capabilities to fabricate very thin plates for CSR cladding, is the main aim of the present study. Making the absorber rod thinner and CSR cladding thicker by introducing a better blackness absorbing material and a new stainless steel alloy, respectively, are two studied ways to reduce the effects of mentioned problems.

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.

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)

Compact torsatron reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Carreras, B.A.; Lynch, V.E.; Tolliver, J.S.; Sviatoslavsky, I.N.

1988-05-01

Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R 0 = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R 0 ≅ 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs

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

Study of reactor parameters of on critical systems, Phase I: Safety report for RB zero power reactor

International Nuclear Information System (INIS)

Raisic, N.

1962-09-01

In addition to the safety analysis for the zero power RB reactor, this report contains a general description of the reactor, reactor components, auxiliary equipment and the reactor building. Reactor Rb has been reconstructed during 1961-1962 and supplied with new safety-control system as well as with a complete dosimetry instrumentation. Since RB reactor was constructed without shielding special attention is devoted to safety and protection of the staff performing experiments. Due to changed circumstances in the Institute ( start-up of the RA 7 MW power reactor) the role of the RB reactor was redefined

Advances in Reactor Physics, Mathematics and Computation. Volume 1

Energy Technology Data Exchange (ETDEWEB)

1987-01-01

These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume one, are divided into 6 sessions bearing on: - session 1: Advances in computational methods including utilization of parallel processing and vectorization (7 conferences) - session 2: Fast, epithermal, reactor physics, calculation, versus measurements (9 conferences) - session 3: New fast and thermal reactor designs (9 conferences) - session 4: Thermal radiation and charged particles transport (7 conferences) - session 5: Super computers (7 conferences) - session 6: Thermal reactor design, validation and operating experience (8 conferences).

Input/output manual of light water reactor fuel analysis code FEMAXI-7 and its related codes

Energy Technology Data Exchange (ETDEWEB)

Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa [Japan Atomic Energy Agency, Nuclear Safety Research Center, Tokai, Ibaraki (Japan); Saitou, Hiroaki [ITOCHU Techno-Solutions Corporation, Tokyo (Japan)

2013-10-15

A light water reactor fuel analysis code FEMAXI-7 has been developed, as an extended version from the former version FEMAXI-6, for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which are fully disclosed in the code model description published in the form of another JAEA-Data/Code report. The present manual, which is the very counterpart of this description document, gives detailed explanations of files and operation method of FEMAXI-7 code and its related codes, methods of input/output, sample Input/Output, methods of source code modification, subroutine structure, and internal variables in a specific manner in order to facilitate users to perform fuel analysis by FEMAXI-7. (author)

Input/output manual of light water reactor fuel analysis code FEMAXI-7 and its related codes

International Nuclear Information System (INIS)

Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa; Saitou, Hiroaki

2013-10-01

A light water reactor fuel analysis code FEMAXI-7 has been developed, as an extended version from the former version FEMAXI-6, for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which are fully disclosed in the code model description published in the form of another JAEA-Data/Code report. The present manual, which is the very counterpart of this description document, gives detailed explanations of files and operation method of FEMAXI-7 code and its related codes, methods of input/output, sample Input/Output, methods of source code modification, subroutine structure, and internal variables in a specific manner in order to facilitate users to perform fuel analysis by FEMAXI-7. (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

Commissioning of research reactors. Safety guide

International Nuclear Information System (INIS)

2006-01-01

The objective of this Safety Guide is to provide recommendations on meeting the requirements for the commissioning of research reactors on the basis of international best practices. Specifically, it provides recommendations on fulfilling the requirements established in paras 6.44 and 7.42-7.50 of International Atomic Energy Agency, Safety of Research Reactors, IAEA Safety Standards Series No. NS-R-4, IAEA, Vienna (2005) and guidance and specific and consequential recommendations relating to the recommendations presented in paras 615-621 of International Atomic Energy Agency, Safety in the Utilization and Modification of Research Reactors, Safety Series No. 35-G2, IAEA, Vienna (1994) and paras 228-229 of International Atomic Energy Agency, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, Safety Series No. 35-G1, IAEA, Vienna (1994). This Safety Guide is intended for use by all organizations involved in commissioning for a research reactor, including the operating organization, the regulatory body and other organizations involved in the research reactor project

Irradiation experiment on fast reactor metal fuels containing minor actinides up to 7 at.% burnup

International Nuclear Information System (INIS)

Ohta, H.; Yokoo, T.; Ogata, T.; Inoue, T.; Ougier, M.; Glatz, J.P.; Fontaine, B.; Breton, L.

2007-01-01

Fast reactor metal fuels containing minor actinides (MAs: Np, Am, Cm) and rare earths (REs) have been irradiated in the fast reactor PHENIX. In this experiment, four types of fuel alloys, U-19Pu-10Zr, U-19Pu-10Zr-2MA-2RE, U-19Pu-10Zr-5MA-5RE and U-19Pu-10Zr-5MA (wt.%), are loaded into part of standard metal fuel stacks. The postirradiation examinations will be conducted at ∼2.4, ∼7 and ∼11 at.% burnup. As for the low-burnup fuel pins, nondestructive postirradiation tests have already been performed and the fuel integrity was confirmed. Furthermore, the irradiation experiment for the intermediate burnup goal of ∼7 at.% was completed in July 2006. For the irradiation period of 356.63 equivalent full-power days, the neutron flux level remained in the range of 3.5-3.6 x 10 15 n/cm 2 /s at the axial peak position. On the other hand, the maximum linear power of fuel alloys decreased gradually from 305-315 W/cm (beginning of irradiation) to 250-260 W/cm (end of irradiation). The discharged peak burnup was estimated to be 6.59-7.23 at.%. The irradiation behavior of MA-containing metal fuels up to 7 at.% burnup was predicted using the ALFUS code, which was developed for U-Pu-Zr ternary fuel performance analysis. As a result, it was evaluated that the fuel temperature is distributed between ∼410 deg. C and ∼645 deg. C at the end of the irradiation experiment. From the stress-strain analysis based on the preliminarily employed cladding irradiation properties and the FCMI stress distribution history, it was predicted that a cladding strain of not more than 0.9% would appear. (authors)

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)

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)

Testing of a 7-tube palladium membrane reactor for potential use in TEP

International Nuclear Information System (INIS)

Carlson, Bryan J.; Trujillo, Stephen; Willms, R. Scott

2010-01-01

A Palladium Membrane Reactor (PMR) consists of a palladium/silver membrane permeator filled with catalyst (catalyst may be inside or outside the membrane tubes). The PMR is designed to recover tritium from the methane, water, and other impurities present in fusion reactor effluent. A key feature of a PMR is that the total hydrogen isotope content of a stream is significantly reduced as (1) methane-steam reforming and/or water-gas shift reactions proceed on the catalyst bed and (2) hydrogen isotopes are removed via permeation through the membrane. With a PMR design matched to processing requirements, nearly complete hydrogen isotope removals can be achieved. A 3-tube PMR study was recently completed. From the results presented in this study, it was possible to conclude that a PMR is appropriate for TEP, perforated metal tube protectors function well, platinum on aluminum (PtA) catalyst performs the best, conditioning with air is probably required to properly condition the Pd/Ag tubes, and that CO/CO 2 ratios maybe an indicator of coking. The 3-tube PMR had a permeator membrane area of 0.0247 m 2 and a catalyst volume to membrane area ratio of 4.63 cc/cm 2 (with the catalyst on the outside of the membrane tubes and the catalyst only covering the membrane tube length). A PMR for TEP will require a larger membrane area (perhaps 0.35 m 2 ). With this in mind, an intermediate sized PMR was constructed. This PMR has 7 permeator tubes and a total membrane area of 0.0851 m 2 . The catalyst volume to membrane area ratio for the 7-tube PMR was 5.18 cc/cm 2 . The total membrane area of the 7-tube PMR (0.0851 m 2 ) is 3.45 times larger than total membrane area of the 3-tube PMR (0.0247 m 2 ). The following objectives were identified for the 7-tube PMR tests: (1) Refine test measurements, especially humidity and flow; (2) Refine maintenance procedures for Pd/Ag tube conditioning; (3) Evaluate baseline PMR operating conditions; (4) Determine PMR scaling method; (5) Evaluate PMR

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

Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 16, 2011, 7:00 PM status

International Nuclear Information System (INIS)

2011-01-01

This situation note is established according to the information gained on March 16, 2011, at 7:00 PM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of all 6 reactors of the Fukushima I site (Dai-ichi) and of their spent fuel pools, as well as the situation of the reactors No. 1, 2, 3 and 4 of the Fukushima II site (Daini), and of the Onagawa and Tokai power plants is briefly presented with the progress of the accident management actions. (J.S.)

Input/output manual of light water reactor fuel performance code FEMAXI-7 and its related codes

Energy Technology Data Exchange (ETDEWEB)

Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa [Japan Atomic Energy Agency, Nuclear Safety Research Center, Tokai, Ibaraki (Japan); Saitou, Hiroaki [ITOCHU Techno-Solutions Corp., Tokyo (Japan)

2012-07-15

A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which has been fully disclosed in the code model description published recently as JAEA-Data/Code 2010-035. The present manual, which is the counterpart of this description, gives detailed explanations of operation method of FEMAXI-7 code and its related codes, methods of Input/Output, methods of source code modification, features of subroutine modules, and internal variables in a specific manner in order to facilitate users to perform a fuel analysis with FEMAXI-7. This report includes some descriptions which are modified from the original contents of JAEA-Data/Code 2010-035. A CD-ROM is attached as an appendix. (author)

Input/output manual of light water reactor fuel performance code FEMAXI-7 and its related codes

International Nuclear Information System (INIS)

Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa; Saitou, Hiroaki

2012-07-01

A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which has been fully disclosed in the code model description published recently as JAEA-Data/Code 2010-035. The present manual, which is the counterpart of this description, gives detailed explanations of operation method of FEMAXI-7 code and its related codes, methods of Input/Output, methods of source code modification, features of subroutine modules, and internal variables in a specific manner in order to facilitate users to perform a fuel analysis with FEMAXI-7. This report includes some descriptions which are modified from the original contents of JAEA-Data/Code 2010-035. A CD-ROM is attached as an appendix. (author)

Fusion reactor materials

International Nuclear Information System (INIS)

Anon.

1977-01-01

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

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

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

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)

European commission - 7th framework programme. The collaborative project on European sodium fast reactor (CP ESFR)

International Nuclear Information System (INIS)

Fiorini, G.L.

2009-01-01

The paper summarizes the key characteristics of the four years large Collaborative Project on European Sodium Fast Reactor (CP ESFR - 2009-2012); the CP ESFR follows the 6th FP project named 'Roadmap for a European Innovative SOdium cooled FAst Reactor - EISOFAR' further identifying, organizing and implementing a significant part of the needed R and D effort. The CP ESFR merges the contribution of 25 european partners; it will be realized under the aegis of the 7th FP under the Area - Advanced Nuclear Systems with a refund from the European Commission of 5.8 M euro (11.55 M euro total budget). It will be a key component of the European Sustainable Nuclear Energy Technology Platform (SNE TP) and its Strategic Research Agenda (SRA). The inputs for the project are the key research goals for fourth generation of European sodium cooled fast reactors which can be summarized as follow: an improved safety with in particular the achievement of a robust architecture vis a vis of abnormal situations and the robustness of the safety demonstrations; the guarantee of a financial risk comparable to that of the other means of energy production; a flexible and robust management of the nuclear materials and especially the waste reduction through the Minor Actinides burning. (author)

76 FR 77855 - Detroit Edison Company; Notice of Availability of Errata Sheet for the Draft Environmental Impact...

Science.gov (United States)

2011-12-14

... Information Contact: Mr. Bruce Olson, Project manager, Environmental Projects Branch 2, Division of New.... Telephone: (301) 415-3731; email: Bruce.Olson@nrc.gov . Dated at Rockville, Maryland, this 7th day of...

Neutronics of a mixed-flow gas-core reactor

International Nuclear Information System (INIS)

Soran, P.D.; Hansen, G.E.

1977-11-01

The study was made to investigate the neutronic feasibility of a mixed-flow gas-core reactor. Three reactor concepts were studied: four- and seven-cell radial reactors and a seven-cell scallop reactor. The reactors were fueled with UF 6 (either U-233 or U-235) and various parameters were varied. A four-cell reactor is not practical nor is the U-235 fueled seven-cell radial reactor; however, the 7-cell U-233 radial and scallop reactors can satisfy all design criteria. The mixed flow gas core reactor is a very attractive reactor concept and warrants further investigation

Light water reactor fuel analysis code FEMAXI-7. Model and structure [Revised edition

International Nuclear Information System (INIS)

Suzuki, Motoe; Udagawa, Yutaka; Amaya, Masaki; Saitou, Hiroaki

2014-03-01

A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in both normal conditions and anticipated transient conditions. This code is an advanced version which has been produced by incorporating the former version FEMAXI-6 with numerous functional improvements and extensions. In FEMAXI-7, many new models have been added and parameters have been clearly arranged. Also, to facilitate effective maintenance and accessibility of the code, modularization of subroutines and functions have been attained, and quality comment descriptions of variables or physical quantities have been incorporated in the source code. With these advancements, the FEMAXI-7 code has been upgraded to a versatile analytical tool for high burnup fuel behavior analyses. This report is the revised edition of the first one which describes in detail the design, basic theory and structure, models and numerical method, and improvements and extensions. The first edition, JAEA-Data/Code 2010-035, was published in 2010. The first edition was extended by orderly addition and disposition of explanations of models and organized as the revised edition after three years interval. (author)

RA Reactor applications, Annex A

International Nuclear Information System (INIS)

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

1990-01-01

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

Compact power reactor

International Nuclear Information System (INIS)

Wetch, J.R.; Dieckamp, H.M.; Wilson, L.A.

1978-01-01

There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector

Molten salt reactors

International Nuclear Information System (INIS)

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

2014-01-01

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

Oklo natural reactor

International Nuclear Information System (INIS)

Fujii, Isao

1985-01-01

In 1954, Professor Kazuo, Kuroda of Arkansas University in USA published the possibility that spontaneously generated natural nuclear reactors existed in prehistoric age. In 1972, 18 years after that, Commissariat a l'Energie Atomique published that in the Oklo uranium deposit in Gabon, Africa, a natural nuclear reactor was found. This fact was immediately informed to the whole world, but in Japan, its details have not necessarily been well known. The chance of investigating into this fact and visiting the Oklo deposit by the favor of COMUF, the owner of the Oklo deposit, was given, therefore, the state of the natural reactors, which has been known so far, is reported. At present, 12 natural reactors have been found in the vicinity of the Oklo deposit. The natural reactors were generated spontaneously in uranium deposits about 1.7 billion years ago when the isotopic abundance of U-235 was 3 %, and the chain reaction started naturally. When the concentration of U-235 lowered, the reaction stopped naturally. The abnormality in the U-235 abundance in natural uranium was found, and the cause was pursued. The evidence of the existence of natural reactors was shown. (Kako, I.)

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)

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.

Enhancing mass transfer and ethanol production in syngas fermentation of Clostridium carboxidivorans P7 through a monolithic biofilm reactor

International Nuclear Information System (INIS)

Shen, Yanwen; Brown, Robert; Wen, Zhiyou

2014-01-01

Highlights: • Syngas fermentation process is limited by gas-to-liquid mass transfer. • A novel monolithic biofilm reactor (MBR) for efficient mass transfer was developed. • MBR with slug flow resulted in higher k L a than bubble column reactor (BCR). • MBR enhanced ethanol productivity by 53% compared to BCR. • MBR was demonstrated as a promising reactor configuration for syngas fermentation. - Abstract: Syngas fermentation is a promising process for producing fuels and chemicals from lignocellulosic biomass. Currently syngas fermentation faces several engineering challenges, with gas-to-liquid mass transfer limitation representing the major bottleneck. The aim of this work is to evaluate the performance of a monolithic biofilm reactor (MBR) as a novel reactor configuration for syngas fermentation. The volumetric mass transfer coefficient (k L a) of the MBR was evaluated in abiotic conditions within a wide range of gas flow rates (i.e., gas velocity in monolithic channels) and liquid flow rates (i.e., liquid velocity in the channels). The k L a values of the MBR were higher than those of a controlled bubble column reactor (BCR) in certain conditions, due to the slug flow pattern in the monolithic channels. A continuous syngas fermentation using Clostridium carboxidivorans P7 was conducted in the MBR system under varying operational conditions, with the variables including syngas flow rate, liquid recirculation between the monolithic column and reservoir, and dilution rate. It was found that the syngas fermentation performance – measured by such parameters as syngas utilization efficiency, ethanol concentration and productivity, and ratio of ethanol to acetic acid – depended not only on the mass transfer efficiency but also on the biofouling or abrading of the biofilm attached on the monolithic channel wall. At a condition of 300 mL/min of syngas flow rate, 500 mL/min of liquid flow rate, and 0.48 day −1 of dilution rate, the MBR produced much higher

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

Annual report, 1981

International Nuclear Information System (INIS)

1982-01-01

In 1981 the government of the province of Ontario announced a major industrial expansion program that gave Ontario Hydro a key role in contributing to the development of the provincial economy. The program looked to an economic structure based increasingly on electricity produced mainly from indigenous fuel sources - uranium and water power - and on nuclear technology. The construction schedule for the Darlington nuclear power station was accelerated as a result. Primary electrical demand reached 101.7 billion kilowatt hours, 36.9 kWh from nuclear sources. An extra 12.5 megagrams of coal would have been naeded to produce this power if nuclear plants had not been available. Of the eight reactors operating at Bruce and Pickering, seven had capacity factors in the top 10 of the world's 130 largest power reactors. Two units at Bruce Heavy Water Plant B were declared in service in 1981. Hydro produced 810 megagrams of heavy water, and bought 970 Mg uranium. The search for alternate energy sources includes active involvement in the fusion energy program and a study of hydrogen supply technology and costs. Employees have been sent to South Korea and Argentina to help commission CANDU nuclear power plants there, and assistance was given to Atomic Energy of Canada Ltd. in preparing a bid for a Mexican station. Radioactive emissions sere cenerally under one percent of the Atomic Energy Control Board limits, and no staff or public injuries or fatalities occurred as a result of the nuclear power program

Mirror reactor studies

International Nuclear Information System (INIS)

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

1977-01-01

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

Calculation of the transmutation rates of Tc-99, I-129 and Cs-135 in the High Flux Reactor, in the Phenix Reactor and in a light water reactor

International Nuclear Information System (INIS)

Bultman, J.

1992-04-01

Transmutation of long-lived fission products is of interest for the reduction of the possible dose to the population resulting from long-term leakage of nuclear waste from waste disposals. Three isotopes are of special interest: Tc-99, I-129 and Cs-135. Therefore, experiments on transmutation of these isotopes in nuclear reactors are planned. In the present study, the possible transmutation rates and mass reductions are determined for experiments in High Flux Reactor (HFR) located in Petten (Netherlands) and in Phenix (France). Also, rates were determined for a standard Light Water Reactor (LWR). The transmutation rates of the 3 fission products will be much higher in HFR than in Phenix reactor, as both total flux and effective cross sections are higher. For thick targets the effective half lives are approximately 3, 2 and 7 years for Tc-99, I-129 and Cs-135 irradiation respectively in HFR and 22, 16 and 40 years for Tc-99, I-129 and Cs-135 irradiation in Phenix reactor. The transmutation rates in LWR are low. Only the relatively large power of LWR guarantees a large total mass reduction. Especially transmutation of Cs-135 will be very difficult in Phenix and LWR, clearly shown by the very long effective half lives of 40 and 100 years, respectively. (author). 7 refs.; 5 figs.; 7 tabs

Structure and thermal properties of as-fabricated U-7Mo/Mg and U-10Mo/Mg low-enriched uranium research reactor fuels

Energy Technology Data Exchange (ETDEWEB)

Kulakov, Mykola, E-mail: mykola.kulakov@cnl.ca [Fuel Development Branch, Canadian Nuclear Laboratories, Chalk River, ON K0J 1J0 Canada (Canada); Saoudi, Mouna [Fuel Development Branch, Canadian Nuclear Laboratories, Chalk River, ON K0J 1J0 Canada (Canada); Piro, Markus H.A. [Fuel and Fuel Channel Safety Branch, Canadian Nuclear Laboratories, Chalk River, ON K0J 1J0 Canada (Canada); Donaberger, Ronald L. [Canadian Neutron Beam Centre, Chalk River, ON K0J 1J0 Canada (Canada)

2017-02-15

Aluminum-clad U-7Mo/Mg and U-10Mo/Mg pin-type mini-elements (with a core uranium loading of 4.5 gU/cm{sup 3}) have been fabricated at the Canadian Nuclear Laboratories for experimental tests and ultimately for use in research and test reactors. In this study, the microstructure and phase composition of unirradiated U-7Mo/Mg and U-10Mo/Mg fuel cores were analyzed using optical and scanning electron microscopy, and neutron powder diffraction. Thermal properties were characterized using a combination of experimental measurements and thermodynamic calculations. The thermal diffusivity was measured using the laser flash method. The temperature-dependent specific heat capacities were calculated based on the linear rule of mixture using the weight fraction of different crystalline phases and their specific heat capacity values taken from the literature. The thermal conductivity was then calculated using the measured thermal diffusivity, the measured density and the calculated specific heat capacity. The resulting thermal conductivity is practically identical for both types of fuel. The in-reactor temperatures were predicted using conjugate heat transfer simulations. - Highlights: • Neutron diffraction analysis shows that most of the γ-U(Mo) phase was retained in as-fabricated U-7Mo/Mg and U-10Mo/Mg fuel cores. • The experimental thermal conductivity of both types of fuel is practically identical. • Based on conjugate heat transfer simulations, under normal operating conditions, the in-reactor fuel centreline temperature is about 510 K.

Progress report concerning safety research for nuclear reactor facilities

International Nuclear Information System (INIS)

1978-01-01

Examination and evaluation of safety research results for nuclear reactor facilities have been performed, as more than a year has elapsed since the plan had been initiated in April, 1976, by the special sub-committee for the safety of nuclear reactor facilities. The research is carried out by being divided roughly into 7 items, and seems to be steadily proceeding, though it does not yet reach the target. The above 7 items include researches for (1) criticality accident, (2) loss of coolant accident, (3) safety for light water reactor fuel, (4) construction safety for reactor facilities, (5) reduction of release of radioactive material, (6) safety evaluation based on the probability theory for reactor facilities, and (7) aseismatic measures for reactor facilities. With discussions on the progress and the results of the research this time, research on the behaviour on fuel in abnormal transients including in-core and out-core experiments has been added to the third item, deleting the power-cooling mismatch experiment in Nuclear Safety Research Reactor of JAERI. Also it has been decided to add two research to the seventh item, namely measured data collection, classification and analysis, and probability assessment of failures due to an earthquake. For these 7 items, the report describes the concrete contents of research to be performed in fiscal years of 1977 and 1978, by discussing on most rational and suitable contents conceivable at present. (Wakatsuki, Y.)

1: the atom. 2: radioactivity. 3: man and radiations. 4: the energy. 5: nuclear energy: fusion and fission. 6: the operation of a nuclear reactor. 7: the nuclear fuel cycle

International Nuclear Information System (INIS)

2002-01-01

This series of 7 digest booklets present the bases of the nuclear physics and of the nuclear energy: 1 - the atom (structure of matter, chemical elements and isotopes, the four fundamental interactions, nuclear physics); 2 - radioactivity (definition, origins of radioelements, applications of radioactivity); 3 - man and radiations (radiations diversity, biological effects, radioprotection, examples of radiation applications); 4 - energy (energy states, different forms of energy, characteristics); 5 - nuclear energy: fusion and fission (nuclear energy release, thermonuclear fusion, nuclear fission and chain reaction); 6 - operation of a nuclear reactor (nuclear fission, reactor components, reactor types); 7 - nuclear fuel cycle (nuclear fuel preparation, fuel consumption, reprocessing, wastes management). (J.S.)

Nuclear Energy Enabling Technologies (NEET) Reactor Materials: News for the Reactor Materials Crosscut, May 2016

Energy Technology Data Exchange (ETDEWEB)

Maloy, Stuart Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Materials Science in Radiation and Dynamics Extremes

2016-09-26

In this newsletter for Nuclear Energy Enabling Technologies (NEET) Reactor Materials, pages 1-3 cover highlights from the DOE-NE (Nuclear Energy) programs, pages 4-6 cover determining the stress-strain response of ion-irradiated metallic materials via spherical nanoindentation, and pages 7-8 cover theoretical approaches to understanding long-term materials behavior in light water reactors.

Fusion reactors as a future energy source

International Nuclear Information System (INIS)

Seifritz, W.

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

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

Simulation development for TRIGA reactor

International Nuclear Information System (INIS)

Handoyo, D.

1997-01-01

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

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

International Nuclear Information System (INIS)

1968-01-01

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

Graphite-water steam-generating reactor in the USSR

Energy Technology Data Exchange (ETDEWEB)

Dollezhal, N A [AN SSSR, Moscow

1981-10-01

One of the types of power reactor used in the USSR is the graphite-water steam-generating reactor RBMK. This produces saturated steam at a pressure of 7MPa. Reactors giving 1GWe each have been installed at the Leningrad, Kursk, Chernobyl and other power stations. Further stations using reactors of this type are being built. A description is given of the fuel element design, and of the layout of the plant. The main characteristics of RBMK reactors using fuel of rated and higher enrichment are listed.

Dose rate in the reactor room and environment during maintenance in fusion reactors

International Nuclear Information System (INIS)

Maki, Koichi; Satoh, Satoshi; Takatsu, Hideyuki; Seki, Yasushi

1995-01-01

According to the International Thermonuclear Experimental Reactor (ITER) conceptual design activity, after reactor shutdown, damaged segments are pulled up from the reactor and hung from the reactor room ceiling by a remote handling device. The dose rate in the reactor room and the environment is estimated for this situation, and the following results are obtained. First, the dose rate in the room is > 10 8 μSv/h. Since this dose rate is 10 7 times greater than the biological radiation shielding design limit of 25 μSv/h, workers cannot enter the room. Second, lenses and optical fiber composed of glass that is radiation resistant up to 10 6 Gy would be damaged after <100 h near the segment, and devices using semiconductors could not work after several hours or so in the aforementioned dose-rate conditions. Third, during suspension of one blanket segment from the ceiling, the dose rate in the site boundary can be reduced by one order by a 23-cm-thicker reactor building roof. To reduce dose rate in public exposure to a value that is less than one-tenth of the public exposure radiation shielding design limit of 100 μSv/yr, the distance of the site boundary from the reactor must be greater than 200 m for a reactor building with a 160-cm-thick concrete roof. 9 refs., 6 figs., 2 tabs

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.)

Expected characteristics of future reactors for human beings

International Nuclear Information System (INIS)

Taketani, Kiyoaki

1992-01-01

Based on four reactor safety components (namely: a) God-given safety, b) Equipment safety, c) Quick-response safety, d) Containing safety), categorical assessment is made of various nuclear reactor concepts ranging from present existing reactors to future reactors based on innovative reactor design. In pursuit of nuclear reactor safety, ultimate characteristics of the ideal nuclear reactor are expected to coincide with those of an inherently safe reactor. A definition of 'inherently safe' has already been proposed by a committee in Japan. As a realistic and existable reactor, which is as close to the ideal reactor, a future reactor which is almost the same as a global reactor, is proposed. This global reactor must be constructable anywhere on earth and must permit easy operation and maintenance by anyone. It is also discussed to identify what behavior is expected of the global reactor under various conditions. At the same time, this future reactor which includes the global reactor, should solve a) the nuclear fuel resource issue, b) efficient utilization of fission energy and c) environmental issues as the greenhouse effect. (author). 7 refs., 2 figs

RA Research reactor Annual report 1982 - Part 1, Operation, maintenance and utilization of the RA reactor

International Nuclear Information System (INIS)

Sotic, O.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.; Miokovic, J.

1982-12-01

Reactor test operation started in September 1981 at 2 MW power with 80% enriched fuel continued during 1982 according to the previous plan. The initial reactor core was made of 44 fuel channel each containing 10 fuel slugs. The first half of 1982 was used for the needed measurements and analysis of operating parameters and functioning of reactor systems and equipment under operating conditions. Program concerned with the testing operation at higher power levels was started in the second half of this year. It was found that the inherent excess reactivity and control rod worths ensure safe operation according to the IAEA safety standards. Excess reactivity is high enough to enable higher power level of 4.7 MW during 4 monthly cycles each lasting 15-20 days. Favourable conditions for cooling exist for the initial core configuration. Effects of poisoning at startup on the reactivity and power density distribution were measured as well as initial spatial distribution of the neutron flux which was 3,9 10 13 cm -2 s -1 at 2 MW power. Modification of the calibration coefficient in the system for automated power level control was determined. All the results show that all the safety criteria and limitations concerned with fuel utilization are fulfilled if reactor power would be 4.7 MW. Additional testing operation at 3, 4, and 4.7 MW power levels will be needed after obtaining the licence for operating at nominal power. Transition from the initial core with 44 fuel channels to the equilibrium lattice configuration with 72 fuel channels each containing 10 fuel slugs, would be done gradually. Reactor was not operated in September because of the secondary coolant pipes were exchanged between Danube and the horizontal sedimentary. Control and maintenance of the reactor equipment was done regularly and efficiently dependent on the availability of the spare parts. Difficulties in maintenance of the reactor instrumentation were caused by unavailability of the outdated spare parts

Practical course on reactor instrumentation

International Nuclear Information System (INIS)

Boeck, H.; Villa, M.

2004-06-01

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

Digital control system of advanced reactor

International Nuclear Information System (INIS)

Peng Huaqing; Zhang Rui; Liu Lixin

2001-01-01

This article produced the Digital Control System For Advanced Reactor made by NPIC. This system uses Siemens SIMATIC PCS 7 process control system and includes five control system: reactor power control system, pressurizer level control system, pressurizer pressure control system, steam generator water level control system and dump control system. This system uses three automatic station to realize the function of five control system. Because the safety requisition of reactor is very strict, the system is redundant. The system configuration uses CFC and SCL. the human-machine interface is configured by Wincc. Finally the system passed the test of simulation by using RETRAN 02 to simulate the control object. The research solved the key technology of digital control system of reactor and will be very helpful for the nationalization of digital reactor control system

Moderator heat recovery of CANDU reactors

International Nuclear Information System (INIS)

Fath, H.E.S.; Ahmed, S.T.

1986-01-01

A moderator heat recovery scheme is proposed for CANDU reactors. The proposed circuit utilizes all the moderator heat to the first stages of the plant feedwater heating system. CANDU-600 reactors are considered with moderator heat load varying from 120 to 160 MWsub(th), and moderator outlet temperature (from calandria) varying from 80 to 100 0 C. The steam saved from the turbine extraction system was found to produce an additional electric power ranging from 5 to 11 MW. This additional power represents a 0.7-1.7% increase in the plant electric output power and a 0.2-0.7% increase in the plant thermal efficiency. The outstanding features and advantages of the proposed scheme are presented. (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

Modeling and performance of the MHTGR [Modular High-Temperature Gas-Cooled Reactor] reactor cavity cooling system

International Nuclear Information System (INIS)

Conklin, J.C.

1990-04-01

The Reactor Cavity Cooling System (RCCS) of the Modular High- Temperature Gas-Cooled Reactor (MHTGR) proposed by the U.S. Department of Energy is designed to remove the nuclear afterheat passively in the event that neither the heat transport system nor the shutdown cooling circulator subsystem is available. A computer dynamic simulation for the physical and mathematical modeling of and RCCS is described here. Two conclusions can be made form computations performed under the assumption of a uniform reactor vessel temperature. First, the heat transferred across the annulus from the reactor vessel and then to ambient conditions is very dependent on the surface emissivities of the reactor vessel and RCCS panels. These emissivities should be periodically checked to ensure the safety function of the RCCS. Second, the heat transfer from the reactor vessel is reduced by a maximum of 10% by the presence of steam at 1 atm in the reactor cavity annulus for an assumed constant in the transmission of radiant energy across the annulus can be expected to result in an increase in the reactor vessel temperature for the MHTGR. Further investigation of participating radiation media, including small particles, in the reactor cavity annulus is warranted. 26 refs., 7 figs., 1 tab

Introduction to reactor internal materials for pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

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

1994-06-01

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

Introduction to reactor internal materials for pressurized water reactor

International Nuclear Information System (INIS)

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

1994-06-01

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

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)

Hanford B Reactor Building Hazard Assessment Report

International Nuclear Information System (INIS)

Griffin, P. W.

1999-01-01

The 105-B Reactor (hereinafter referred to as B Reactor) is located in the 100 Area of the Hanford Site near Richland, Washington. The B Reactor is one of nine plutonium production reactors that were constructed in the 1940s during the Cold War Era. Construction of the B Reactor began June 7, 1943, and operation began on September 26, 1944. The Environmental Restoration Contractor was requested by RL to provide an assessment/characterization of the B Reactor building to determine and document the hazards that are present and could pose a threat to the environment and/or to individuals touring the building. This report documents the potential hazards, determines the feasibility of mitigating the hazards, and makes recommendations regarding areas where public tour access should not be permitted

Results and recommendations from the reactor chemistry and corrosion tasks of the reactor materials program

International Nuclear Information System (INIS)

Baumann, E.W.; Ondrejcin, R.S.

1990-11-01

Within the general context of extended service life, the Reactor Materials Program was initiated in 1984. This comprehensive program addressed material performance in SRS reactor tanks and the primary coolant or Process Water System (PWS) piping. Three of the eleven tasks concerned moderator quality and corrosion mitigation. Definition and control of the stainless steel aqueous environment is a key factor in corrosion mitigation. The Reactor Materials Program systematically investigated the SRS environment and its effect on crack initiation and propagation in stainless steel, with the objective of improving this environment. The purpose of this report is to summarize the contributions of Tasks 6, 7 and 10 of the Reactor Materials Program to the understanding and control of moderator quality and its relationship to mitigation of stress corrosion cracking

University Reactor Instrumentation grant program. Final report, September 7, 1990--August 31, 1995

International Nuclear Information System (INIS)

Talnagi, J.W.

1998-01-01

The Ohio State University Nuclear Reactor Laboratory (OSU NRL) participated in the Department of Energy (DOE) grant program commonly denoted as the University Reactor Instrumentation (URI) program from the period September 1990 through August 1995, after which funding was terminated on a programmatic basis by DOE. This program provided funding support for acquisition of capital equipment targeted for facility upgrades and improvements, including modernizing reactor systems and instrumentation, improvements in research and instructional capabilities, and infrastructure enhancements. The staff of the OSU NRL submitted five grant applications during this period, all of which were funded either partially or in their entirety. This report will provide an overview of the activities carried out under these grants and assess their impact on the OSU NRL facilities

The STAT7 Code for Statistical Propagation of Uncertainties In Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

Energy Technology Data Exchange (ETDEWEB)

Dunn, Floyd E. [Argonne National Lab. (ANL), Argonne, IL (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States). Nuclear Reactor Lab.; Wilson, Erik [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-12-01

The STAT code was written to automate many of the steady-state thermal hydraulic safety calculations for the MIT research reactor, both for conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel and for future fuel re-loads after the conversion. A Monte-Carlo statistical propagation approach is used to treat uncertainties in important parameters in the analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings on avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis has been performed by analyzing the margin to ONB for each axial node of each stripe of each plate of each element in the core.

FISA 2009 - 7th European Commission conference on EURATOM research and training in reactor systems. Conference proceedings

International Nuclear Information System (INIS)

Goethem, G. van; Manolatos, P.; Hugon, M.; Bhatnagar, V.; Deffrennes, M.; Webster, S.

2010-01-01

The main achievements of the first series of projects under EURATOM FP-7 for nuclear research and training activities (2007 to 2011) were discussed. Approximately 500 participants were registered at FISA 2009 and at the 7 post-conference workshops, representing a wide audience of nuclear scientists and decision makers coming from 32 countries worldwide. The focus of the conference was on scientific and technological research in the following areas: nuclear plant life management for existing reactors (Generation II), severe accident management (Generation III), assessment of future nuclear fission systems (Generation IV), partitioning and transmutation systems (innovative fuels), access to large research infrastructures, and nuclear education and training. Special attention was devoted to the societal and industrial goals of GIF: sustainability, industrial competitiveness, safety and reliability, proliferation resistance. (orig.)

MLR reactor

International Nuclear Information System (INIS)

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

1998-01-01

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

Bruce B fuelling-with-flow operations: fuel damage investigation

Energy Technology Data Exchange (ETDEWEB)

Manzer, A.M. [CANTECH Associates Ltd., Burlington, Ontario (Canada); Morikawa, D. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Hains, A.J.; Cichowlas, W.M. [Nuclear Safety Solutions Limited, Toronto, Ontario (Canada); Roberts, J.G.; Wylie, J. [Bruce Power, Ontario (Canada)

2005-07-01

This paper summarizes the fuel bundle damage characterization done by Nuclear Safety Solutions Limited (NSS) and the out-reactor flow visualization tests done at Atomic Energy of Canada Limited (AECL) to reproduce the damage observed on irradiated fuel bundles. The bearing pad damage mechanism was identified and the tests showed that a minor change to the fuelling sequence would eliminate the mechanical interaction. The change was implemented in January 2005. Since then, the bearing pad damage appears to have been greatly reduced based on the small number of discharged bundles inspected to date. (author)

Bruce B fuelling-with-flow operations: fuel damage investigation

International Nuclear Information System (INIS)

Manzer, A.M.; Morikawa, D.; Hains, A.J.; Cichowlas, W.M.; Roberts, J.G.; Wylie, J.

2005-01-01

This paper summarizes the fuel bundle damage characterization done by Nuclear Safety Solutions Limited (NSS) and the out-reactor flow visualization tests done at Atomic Energy of Canada Limited (AECL) to reproduce the damage observed on irradiated fuel bundles. The bearing pad damage mechanism was identified and the tests showed that a minor change to the fuelling sequence would eliminate the mechanical interaction. The change was implemented in January 2005. Since then, the bearing pad damage appears to have been greatly reduced based on the small number of discharged bundles inspected to date. (author)

The SPHINX reactor for engineering tests

International Nuclear Information System (INIS)

Adamov, E.O.; Artamkin, K.N.; Bovin, A.P.; Bulkin, Y.M.; Kartashev, E.F.; Korneev, A.A.; Stenbok, I.A.; Terekhov, A.S.; Khmel'Shehikov, V.V.; Cherkashov, Y.M.

1990-01-01

A research reactor known as SPHINX is under development in the USSR. The reactor will be used mainly to carry out tests on mock-up power reactor fuel assemblies under close-to-normal parameters in experimental loop channels installed in the core and reflector of the reactor, as well as to test samples of structural materials in ampoule and loop channels. The SPHINX reactor is a channel-type reactor with light-water coolant and moderator. Maximum achievable neutron flux density in the experimental channels (cell composition 50% Fe, 50% H 2 O) is 1.1 X 10 15 neutrons/cm 2 · s for fast neutrons (E > 0.1 MeV) and 1.7 X 10 15 for thermal neutrons at a reactor power of 200 MW. The design concepts used represent a further development of the technical features which have met with approval in the MR and MIR channel-type engineering test reactors currently in use in the USSR. The 'in-pond channel' construction makes the facility flexible and eases the carrying out of experimental work while keeping discharges of radioactivity into the environment to a low level. The reactor and all associated buildings and constructions conform to modern radiation safety and environmental protection requirements

Submersion-Subcritical Safe Space (S4) reactor

International Nuclear Information System (INIS)

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

2006-01-01

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

Proceedings of the 1. international conference on CANDU fuel handling systems

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-12-31

Besides information on fuel loading and handling systems for CANDU and PHWR reactors, the 25 papers in these proceedings also include some on dry storage, modification to fuel strings at Bruce A, and on the SLAR (spacer location and repositioning) system for finding and moving garter springs. The individual papers have been abstracted separately.

Proceedings of the 1. international conference on CANDU fuel handling systems

International Nuclear Information System (INIS)

1996-01-01

Besides information on fuel loading and handling systems for CANDU and PHWR reactors, the 25 papers in these proceedings also include some on dry storage, modification to fuel strings at Bruce A, and on the SLAR (spacer location and repositioning) system for finding and moving garter springs. The individual papers have been abstracted separately

Very-high-temperature reactors for future use

International Nuclear Information System (INIS)

Kasten, P.R.

1988-08-01

Very-high-temperature reactors (VHTRs) show promise for economic generation of electricity and of high-temperature process heat. The key is the development of high-temperature materials which permit gas turbine VHTRs to generate electricity economically, at reactor coolant temperatures which can be used for fossil fuel conversion processes. 7 refs., 5 figs

RAPK-7. code for calculating mass transfer and corrosion products activation in the circulation loops of water-cooled reactors

International Nuclear Information System (INIS)

Mikhaylov, A.V.; Moryakov, A.V.; Nikitin, A.V.

2012-09-01

The RAPK-7 code was developed to simulate formation of non-irradiated and activated corrosion products, their transport and deposition on inner surfaces of primary components and in primary coolant of water-cooled reactors during their operation on power and after shutdown. The key feature of this code is its particular emphasis on the contamination of circulation loops by radioactive corrosion products of reactor which operates on variable modes. Such reactors typically are: research reactors and their experimental loops, naval nuclear power systems, etc. It's typical for such reactors to have repeated (over the campaign) and frequent variations in power (activating neutron fluxes), thermal-physical, hydrodynamic and other parameters of coolant, intensive water mass exchange between the circulation loop and the pressuriser, etc. The processes of mass-transfer are described by the RAPK-7 code with the use of models similar to those employed by the COTRAN and PACTOLE codes. The circulation circuit is broken down into computation areas. The user will then set the concentrations of water chemistry adjusting additives (alkali, boric acid, ammonia, hydrogen), as well as parameters in each area, such as wall temperature, coolant flow core temperature, pressure, flow rate, velocity, the radial component of coolant flowrate and activating neutron flux density. All the above parameters can be set as time-dependent step functions (bar charts), with independent time steps for each of them. The number of computation areas, the number of time dependencies and the level of detail in their description are limited by computer capabilities only. A 'brake' mode with a single-step change of the required set of parameters is provided to allow for jump-type events, such as replacement of contaminated components with clean ones during core refueling or repairs, emergency injection of boric acid, water mass exchange between the circulation circuit and the pressuriser, etc

Behavior of low-burnup metallic fuels for the integral fast reactor at elevated temperatures in ex-reactor tests

International Nuclear Information System (INIS)

Tsai, Hanchung; Liu, Yung Y.; Wang, Da-Yung; Kramer, J.M.

1991-07-01

A series of ex-reactor heating tests on low burnup U-26wt.%Pu-10wt.%Zr metallic fuel for the PRISM reactor was conducted to evaluate fuel/cladding metallurgical interaction and its effect on cladding integrity at elevated temperatures. The reaction between the fuel and cladding caused liquid-phase formation and dissolution of the inner surface of the cladding. The rate of cladding penetration was below the existing design correlation, which provides a conservative margin to cladding failure. In a test which enveloped a wide range of postulated reactor transient events, a substantial temporal cladding integrity margin was demonstrated for an intact, whole fuel pin. The cause of the eventual pin breach was reaction-induced cladding thinning combined with fission-gas pressure loading. The behavior of the breached pin was benign. 7 refs., 7 figs., 1 tab

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

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

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

Global risk of radioactive fallout after major nuclear reactor accidents

International Nuclear Information System (INIS)

Lelieveld, J.; Kunkel, D.; Lawrence, M.G.

2012-01-01

Major reactor accidents of nuclear power plants are rare, yet the consequences are catastrophic. But what is meant by ''rare''? And what can be learned from the Chernobyl and Fukushima incidents? Here we assess the cumulative, global risk of exposure to radioactivity due to atmospheric dispersion of gases and particles following severe nuclear accidents (the most severe ones on the International Nuclear Event Scale, INES 7), using particulate "1"3"7Cs and gaseous "1"3"1I as proxies for the fallout. Our results indicate that previously the occurrence of INES 7 major accidents and the risks of radioactive contamination have been underestimated. Using a global model of the atmosphere we compute that on average, in the event of a major reactor accident of any nuclear power plant worldwide, more than 90% of emitted "1"3"7Cs would be transported beyond 50 km and about 50% beyond 1000 km distance before being deposited. This corroborates that such accidents have large-scale and trans-boundary impacts. Although the emission strengths and atmospheric removal processes of "1"3"7Cs and "1"3"1I are quite different, the radioactive contamination patterns over land and the human exposure due to deposition are computed to be similar. High human exposure risks occur around reactors in densely populated regions, notably in West Europe and South Asia, where a major reactor accident can subject around 30 million people to radioactive contamination. The recent decision by Germany to phase out its nuclear reactors will reduce the national risk, though a large risk will still remain from the reactors in neighbouring countries.

What next for Fukushima?

Science.gov (United States)

Drinkwater, Bruce; Malkin, Rob

2018-01-01

Nearly seven years after a powerful tsunami caused catastrophic damage to Japan's Fukushima Daiichi nuclear-power plant, the clean-up and recovery is still ongoing. Bruce Drinkwater and Rob Malkin recently visited the disaster site and the undamaged Tsuruga plant to see if they can pinpoint the true extent of the damage in the dangerously radioactive reactors

Assessment of torsatrons as reactors

International Nuclear Information System (INIS)

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

1992-12-01

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

Reactor core in FBR type reactor

International Nuclear Information System (INIS)

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

1989-01-01

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

Annual report on the state of RB reactor components and equipment, december 1998

International Nuclear Information System (INIS)

Milosevic, M.

1998-12-01

According to the performed analysis, it is considered that the RB reactor can be operated safely until the existing control and safety systems could be maintained in satisfactory operable state. Failures of heavy water circulation system valves which may cause decreased availability but no accident. During 1998 the reactor lattice was changed 7 times, meaning that experiments were done with 7 configurations of the reactor core. Total reactor operation amounted to 177.5 Wh with 40 start-ups (attained criticality levels). This report contains 4 Annexes, detailed description of the state of reactor equipment in 1998, reactor operation nd utilization data, plan for regular annual maintenance and refurbishment of reactor equipment and plan for minimum needed resources for regular maintenance of the components and equipment in the forthcoming year

BNL ALARA Center: ALARA Notes, No. 9

International Nuclear Information System (INIS)

Khan, T.A.; Xie, J.W.; Beckman, M.C.

1994-02-01

This issue of the Brookhaven National Laboratory's Alara Notes includes the agenda for the Third International Workshop on ALARA and specific instructions on the use of the on-line fax-on-demand service provided by BNL. Other topics included in this issue are: (1) A discussion of low-level discharges from Canadian nuclear plants, (2) Safety issues at French nuclear plants, (3) Acoustic emission as a means of leak detection, (4) Replacement of steam generators at Doel-3, Beaznau, and North Anna-1, (5) Remote handling equipment at Bruce, (6) EPRI's low level waste program, (7) Radiation protection during concrete repairs at Savannah River, (8) Reactor vessel stud removal/repair at Comanche Peak-1, (9) Rework of reactor coolant pump motors, (10) Restoration of service water at North Anna-1 and -2, (11) Steam generator tubing problems at Mihama-1, (12) Full system decontamination at Indian Point-2, (13) Chemical decontamination at Browns Ferry-2, and (14) Inspection methodolody in France and Japan

Effluent releases at the TRIGA reactor facility

Energy Technology Data Exchange (ETDEWEB)

Whittemore, W L [General Atomic Co., San Diego, CA (United States)

1974-07-01

The principal effluent from the operating TRIGA reactors in our facility is argon-41. As monitored by a recording gas and particulate stack monitor, the values shown in the table, the Mark III operating 24 hours per day for very long periods produced the largest amount of radioactive argon. The quantity of 23.7 Ci A-41 when diluted by the normal reactor room ventilation system corresponded to 1.45 x 10{sup -6} {mu}Ci/cc. As diluted in the roof stack stream and the reactor building wake, the concentration immediately outside the reactor building was 25% MPC for an unrestricted area. The continued dilution of this effluent resulted in a concentration of a few percent MPC at the site boundary (unrestricted area) 350 meters from the reactor. (author)

Target-fueled nuclear reactor for medical isotope production

Science.gov (United States)

Coats, Richard L.; Parma, Edward J.

2017-06-27

A small, low-enriched, passively safe, low-power nuclear reactor comprises a core of target and fuel pins that can be processed to produce the medical isotope .sup.99Mo and other fission product isotopes. The fuel for the reactor and the targets for the .sup.99Mo production are the same. The fuel can be low enriched uranium oxide, enriched to less than 20% .sup.235U. The reactor power level can be 1 to 2 MW. The reactor is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7 to 21 days.

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

Energy Technology Data Exchange (ETDEWEB)

Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Mueller, Donald E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

2016-08-31

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 ⁷LiF-BeF₂) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

International Nuclear Information System (INIS)

Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.; Powers, Jeffrey J.

2016-01-01

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 "7LiF-BeF_2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

A two dimensional code (R,Z) for nuclear reactor analysis and its application to the UAR-RI reactor

International Nuclear Information System (INIS)

Bishay, S.T.; Mikhail, I.F.I.; Gaafar, M.A.; Michaiel, M.L.; Nassar, S.F.

1988-01-01

A detailed study is given of fuel consumption in completely reflected cylindrical reactors. A two group, two dimensional (r,z) code is developed to carry out the required procedure. The code is applied to the UAR-RI reactor and the results are found to be in complete agreement with the experimental observations and with the theoretical interpretations. 7 fig., 12 tab

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Lesson Learned in Preparation for Decommissioning of Three Canadian Prototype Power Reactors

International Nuclear Information System (INIS)

Vickerd, Meggan; Kenny, Stephen

2016-01-01

Lesson learned by Canadian Nuclear Laboratories (CNL)(former AECL) in preparation for decommissioning of three Prototype Reactors is a result of various strategies used for each site. CNL is responsible for the eventual decommissioning of three prototype power reactors; Nuclear Power Demonstration (NPD), Gentilly-1 and Douglas Point. Each of the Canadian prototype power reactor sites shutdown using different strategies. Depending on the site location, configuration, and intended designation of the respective sites, the individual facility systems (ventilation, electrical system, fire detection etc.) were also shut down using different strategies and operating objectives. As CNL embarks on decommissioning the first Canadian prototype reactor, this paper will reflect on the lessons learned over the past thirty years and what CNL is adjusting in the decommissioning strategy to prepare better plans for the future. The Nuclear Power Demonstration Nuclear Generating Station (NPDNGS) was constructed in late 1950's and operated from 1962 to 1987 when it was permanently shutdown after exceeding its operational goals. The NPD reactor was the first Canadian nuclear power reactor and it consisted of a single 20 MWe pressurized heavy water reactor located on a single facility site in Rolphton, Ontario. The NPD facility was shutdown to a 'Cold, Dark and Quiet' state and is maintained using an unmanned strategy by managing the site remotely with active fire detection and security surveillance systems, minimal electrical supply and an active ventilation system which is operated periodically to allow for intermittent inspections. The Douglas Point Nuclear Generating Station (DPNGS) was constructed in the early 1960's and operated from 1968 to 1984 when it was permanently shutdown. It consisted of a 200 MW prototype Canada Deuterium Uranium (CANDU) reactor and is embedded on the Bruce Power site near Kincardine, Ontario. The Douglas Point site is maintained in a

Nuclear reactor refueling system

International Nuclear Information System (INIS)

Wade, E.E.

1978-01-01

A system for transferring fuel assemblies between a nuclear reactor core and a fuel storage area while the fuel assembies remain completely submerged in a continuous body of coolant is described. The system comprises an in-vessel fuel transfer machine located inside the reactor vessel and an ex-vessel fuel transfer machine located in a fuel storage tank. The in-vessel fuel transfer machine comprises two independently rotatable frames with a pivotable fuel transfer apparatus disposed on the lower rotatable frame. The ex-vessel fuel transfer machine comprises one frame with a pivotable fuel transfer apparatus disposed thereon. The pivotable apparatuses are capable of being aligned with each other to transfer a fuel assembly between the reactor vessel and fuel storage tank while the fuel assembly remains completely submerged in a continuous body of coolant. 9 claims, 7 figures

Remote handling equipment for CANDU retubing

International Nuclear Information System (INIS)

Crawford, G.S.; Lowe, H.

1993-01-01

Numet Engineering Ltd. has designed and supplied remote handling equipment for Ontario Hydro's retubing operation of its CANDU reactors at the Bruce Nuclear Generating Station. This equipment consists of ''Retubing Tool Carriers'' an'' Worktables'' which operate remotely or manually at the reactor face. Together they function to transport tooling to and from the reactor face, to position and support tooling during retubing operations, and to deliver and retrieve fuel channels and channel components. This paper presents the fundamentals of the process and discusses the equipment supplied in terms of its design, manufacturing, components and controls, to meet the functional and quality requirements of Ontario Hydro's retubing process. (author)

Load-following performance and assessment of CANDU fuel

Energy Technology Data Exchange (ETDEWEB)

Tayal, M.; Floyd, M.; Rattan, D.; Xu, Z.; Manzer, A.; Lau, J. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Kohn, E. [Ontario Power Generation, Fuel and Fuel Channel Analysis Dept., Toronto, Ontario (Canada)

1999-09-01

Load following of nuclear reactors is now becoming an economic necessity in some countries. When nuclear power stations are operated in a load-following mode, the reactor and the fuel may be subjected to step changes in power on a weekly, daily, or even hourly basis, depending on the grid's needs. This paper updates the previous surveys of load-following capability of CANDU fuel, focusing mainly on the successful experience at the Bruce B station. As well, initial analytical assessments are provided that illustrate the capability of CANDU fuel to survive conditions other than those for which direct in-reactor evidence is available. (author)

Reactor core and initially loaded reactor core of nuclear reactor

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo.

1989-01-01

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

“Wer jetzt kein Haus hat, baut sich keines mehr”. Notes on Bruce O’Neill’s "The Space of Boredom"

Directory of Open Access Journals (Sweden)

George Tudorie

2017-12-01

Full Text Available Bruce O’Neill describes homeless men and women in Bucharest who could not navigate the downward spiral which followed the meltdown of the state-run economy after 1989. Deprived of the culturally treasured anchors of work, home, and family, and unable to participate in anything recognizably meaningful, these individuals are forced into a position of malignant contemplation, even when busy surviving. It is an experience of paralyzed restlessness which resonates with the ruins they zigzag through. The Space of Boredom captures this landscape convincingly, and in elegant prose. The book moves effortlessly from the discussion of scholarly works in a number of fields, to observation of sometimes cinematic quality. It is well argued, abundantly researched, and clear about its theoretical assumptions. If some questions remain to be answered, at least to this reader, this may be an artifact of background. Assumptions, including important ones about the nature of affects, and about boredom itself, may not be shared. Some questions of this kind will be raised in the following.

Neutron field control cybernetics model of RBMK reactor operator

International Nuclear Information System (INIS)

Polyakov, V.V.; Postnikov, V.V.; Sviridenkov, A.N.

1992-01-01

Results on parameter optimization for cybernetics model of RBMK reactor operator by power release control function are presented. Convolutions of various criteria applied previously in algorithms of the program 'Adviser to reactor operator' formed the basis of the model. 7 refs.; 4 figs

Some novel on-power refuelling features of CANDU stations

International Nuclear Information System (INIS)

Erwin, D.; Pendlebury, B.; Watson, J.F.; Welch, A.C.

1976-01-01

Part A of the paper describes the reasons for, and advantages resulting from, the use of flow assisted refuelling in the CANDU type nuclear reactors at the Pickering Generating Station. A separate fuel handling system is used for each reactor unit, as distinct from the system employed at the Bruce Generating station, where the fuel handling system is shared among several units. Part B of the paper describes some of the advantages of the shared concept with particular emphasis on the availability of the fuel handling system. (author)

Economic incentives of short out-of-reactor time for fast breeder reactor fuel

International Nuclear Information System (INIS)

Bentley, B.W.; Haffner, D.R.

1975-01-01

Economic benefits (primarily reduced uranium ore and enrichment expenditures) can be realized by reducing the LMFBR out-of-reactor fuel cycle time only if process cost penalties and R and D costs can be minimized. The results of the evaluation presented show the potential gross benefits of reducing the out-of-reactor time and the effects of various associated cost penalties on these benefits. The gross benefit results estimate the potential savings in electrical power generation in the next 50 years using constant 1975 dollars and discounting the costs at 7 1 / 2 percent per year

Gas turbine modular helium reactor in cogeneration; Turbina de gas reactor modular con helio en cogeneracion

Energy Technology Data Exchange (ETDEWEB)

Leon de los Santos, G. [UNAM, Facultad de Ingenieria, Division de Ingenieria Electrica, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico, D. F. (Mexico)], e-mail: tesgleon@gmail.com

2009-10-15

This work carries out the thermal evaluation from the conversion of nuclear energy to electric power and process heat, through to implement an outline gas turbine modular helium reactor in cogeneration. Modeling and simulating with software Thermo flex of Thermo flow the performance parameters, based on a nuclear power plant constituted by an helium cooled reactor and helium gas turbine with three compression stages, two of inter cooling and one regeneration stage; more four heat recovery process, generating two pressure levels of overheat vapor, a pressure level of saturated vapor and one of hot water, with energetic characteristics to be able to give supply to a very wide gamma of industrial processes. Obtaining a relationship heat electricity of 0.52 and efficiency of net cogeneration of 54.28%, 70.2 MW net electric, 36.6 MW net thermal with 35% of condensed return to 30 C; for a supplied power by reactor of 196.7 MW; and with conditions in advanced gas turbine of 850 C and 7.06 Mpa, assembly in a shaft, inter cooling and heat recovery in cogeneration. (Author)

The safety basis of the integral fast reactor program

International Nuclear Information System (INIS)

Pedersen, D.R.; Seidel, B.R.

1990-01-01

The Integral Fast Reactor (IFR) and metallic fuel have emerged as the US Department of Energy reference reactor concept and fuel system for the development of an advanced liquid-metal reactor. This article addresses the basic elements of the IFR reactor concept and focuses on the safety advances achieved by the IFR Program in the areas of (1) fuel performance, (2) superior local faults tolerance, (3) transient fuel performance, (4) fuel-failure mechanisms, (5) performance in anticipated transients without scram, (6) core-melt mitigation, and (7) actinide recycle

RELAP-7 Software Verification and Validation Plan

Energy Technology Data Exchange (ETDEWEB)

Smith, Curtis L. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk, Reliability, and Regulatory Support; Choi, Yong-Joon [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk, Reliability, and Regulatory Support; Zou, Ling [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk, Reliability, and Regulatory Support

2014-09-25

This INL plan comprehensively describes the software for RELAP-7 and documents the software, interface, and software design requirements for the application. The plan also describes the testing-based software verification and validation (SV&V) process—a set of specially designed software models used to test RELAP-7. The RELAP-7 (Reactor Excursion and Leak Analysis Program) code is a nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on the INL’s modern scientific software development framework – MOOSE (Multi-Physics Object-Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5’s capability and extends the analysis capability for all reactor system simulation scenarios.

Nuclear reactor physics course for reactor operators

International Nuclear Information System (INIS)

Baeten, P.

2006-01-01

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

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

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

Reactor physics calculations in the Nordic countries

International Nuclear Information System (INIS)

Hoeglund, R.

1995-01-01

The seventh biennial meeting on reactor physics calculations in the Nordic countries was arranged by VTT Energy on May 8-9, 1995. 26 papers on different subjects in the field of reactor physics were presented by 45 participants representing research establishments, technical universities, utilities, consultants and suppliers. Resent development and verification of the program systems of ABB Atom, Risoe, Scandpower, Studsvik and VTT Energy were the main topic of the meeting. Benchmarking of the two assembly codes CASMO-4 and HELIOS is proceeding. Cross section data calculated with CASMO-HEX have been validated for the Loviisa reactors. On core analysis ABB atom gives a description on its latest core simulator version POLCA7 with the calculation Core Master 2 and the BWR core supervision system Core Watch. Transient calculations with HEXTRAN, HEXTRAN- PLIM, TRAB, RAMONA, SIMULATE-3K and a code based on PRESTO II/POLCA7 were also presented

RELAP-7 Closure Correlations

Energy Technology Data Exchange (ETDEWEB)

Zou, Ling [Idaho National Lab. (INL), Idaho Falls, ID (United States); Berry, R. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Martineau, R. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Andrs, D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, H. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hansel, J. E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sharpe, J. P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Johns, Russell C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-04-01

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The code is based on the INL’s modern scientific software development framework, MOOSE (Multi-Physics Object Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5’s and TRACE’s capabilities and extends their analysis capabilities for all reactor system simulation scenarios. The RELAP-7 code utilizes the well-posed 7-equation two-phase flow model for compressible two-phase flow. Closure models used in the TRACE code has been reviewed and selected to reflect the progress made during the past decades and provide a basis for the colure correlations implemented in the RELAP-7 code. This document provides a summary on the closure correlations that are currently implemented in the RELAP-7 code. The closure correlations include sub-grid models that describe interactions between the fluids and the flow channel, and interactions between the two phases.

Nuclear reactors

International Nuclear Information System (INIS)

Barre, Bertrand

2015-10-01

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

Fusion reactor safety

International Nuclear Information System (INIS)

1987-12-01

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

Inherent safety characteristics of innovative reactors

International Nuclear Information System (INIS)

Heil, J.A.

1995-11-01

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

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.)

Innovative energy production in fusion reactors

International Nuclear Information System (INIS)

Iiyoshi, A.; Momota, H.; Motojima, O.

1994-01-01

Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are (a) traveling wave direct energy conversion of 14.7 MeV protons, (b) cusp type direct energy conversion of charged particles, (c) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas, and (d) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising. (author)

Innovative energy production in fusion reactors

International Nuclear Information System (INIS)

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

1993-10-01

Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are (a) traveling wave direct energy conversion of 14.7 MeV protons, (b) cusp type direct energy conversion of charged particles, (c) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas, and (d) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising. (author)

Innovative energy production in fusion reactors

Science.gov (United States)

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

1993-10-01

Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are: (1) traveling wave direct energy conversion of 14.7 MeV protons; (2) cusp type direct energy conversion of charged particles; (3) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas; and (4) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising.

Innovative energy production in fusion reactors

Energy Technology Data Exchange (ETDEWEB)

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

1993-10-01

Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are (a) traveling wave direct energy conversion of 14.7 MeV protons, (b) cusp type direct energy conversion of charged particles, (c) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas, and (d) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising. (author).

Effective utilization and management of research reactors

Energy Technology Data Exchange (ETDEWEB)

Muranaka, R [International Atomic Energy Agency, Vienna (Austria). Div. of Research and Isotopes

1984-06-01

The problem of utilizing a research reactor effectively is closely related to its management and therefore should not be considered separately. Too often, attention has been focused on specific techniques and methods rather than on the overall programme of utilization, with the result that skills and equipment have been acquired without any active continuing programme of applications and services. The seminar reported here provided a forum for reactor managers, users, and operators to discuss their experience. At the invitation of the Government of Malaysia, it was held at the Asia Pacific Development Centre, Kuala Lumpur, from 7 to 11 November 1983. It was attended by about 50 participants from 19 Member States; it is hoped that a report on the seminar, including papers presented, can be published and thus reach a wider audience. Thirty-one lectures and contributions were presented at a total of seven sessions: Research reactor management; Radiation exposure and safety; Research reactor utilization (two sessions); PUSPATI Research Reactor Project Development; Core conversion to low-enriched uranium, and safeguards; Research reactor technology. In addition, a panel discussed the causes and resolutions of the under-utilization of research reactors.

Effective utilization and management of research reactors

International Nuclear Information System (INIS)

Muranaka, R.

1984-01-01

The problem of utilizing a research reactor effectively is closely related to its management and therefore should not be considered separately. Too often, attention has been focused on specific techniques and methods rather than on the overall programme of utilization, with the result that skills and equipment have been acquired without any active continuing programme of applications and services. The seminar reported here provided a forum for reactor managers, users, and operators to discuss their experience. At the invitation of the Government of Malaysia, it was held at the Asia Pacific Development Centre, Kuala Lumpur, from 7 to 11 November 1983. It was attended by about 50 participants from 19 Member States; it is hoped that a report on the seminar, including papers presented, can be published and thus reach a wider audience. Thirty-one lectures and contributions were presented at a total of seven sessions: Research reactor management; Radiation exposure and safety; Research reactor utilization (two sessions); PUSPATI Research Reactor Project Development; Core conversion to low-enriched uranium, and safeguards; Research reactor technology. In addition, a panel discussed the causes and resolutions of the under-utilization of research reactors

Processes influencing cooling of reactor effluents

International Nuclear Information System (INIS)

Magoulas, V.E.; Murphy, C.E. Jr.

1982-01-01

Discharge of heated reactor cooling water from SRP reactors to the Savannah River is through sections of stream channels into the Savannah River Swamp and from the swamp into the river. Significant cooling of the reactor effluents takes place in both the streams and swamp. The majority of the cooling is through processes taking place at the surface of the water. The major means of heat dissipation are convective transfer of heat to the air, latent heat transfer through evaporation and radiative transfer of infrared radiation. A model was developed which incorporates the effects of these processes on stream and swamp cooling of reactor effluents. The model was used to simulate the effect of modifications in the stream environment on the temperature of water flowing into the river. Environmental effects simulated were the effect of changing radiant heat load, the effect of changes in tree canopy density in the swamp, the effect of total removal of trees from the swamp, and the effect of diverting the heated water from L reactor from Steel Creek to Pen Branch. 6 references, 7 figures

David Schimmelpenninck van der Oye and Bruce W. Menning, eds., Reforming the Tsar's Army: Military Innovation in Imperial Russia from Peter the Great to the Revolution, Washington, DC: Woodrow Wilson Center and Cambridge: Cambridge UP, 2004.

Directory of Open Access Journals (Sweden)

David Stone

2006-11-01

Full Text Available David Schimmelpenninck and Bruce Menning have produced an excellent volume collecting contributions of a number of both well-established and junior scholars on the history and development of the tsarist military, grouped together around the general theme of reform. In some ways, it is comparable to Eric Lohr and Marshall Poe's complementary The Military and Society in Russia, 1450-1917 (Leiden, 2002. Schimmelpenninck and Menning's contributors, however, focus more on political and institutio...

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

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

Nuclear process steam for industry

International Nuclear Information System (INIS)

Seddon, W.A.

1981-11-01

A joint industrial survey funded by the Bruce County Council, the Ontario Energy Corporation and Atomic Energy of Canada Limited was carried out with the cooperation of Ontario Hydro and the Ontario Ministry of Industry and Tourism. Its objective was to identify and assess the future needs and interest of energy-intensive industries in an Industrial Energy Park adjacent to the Bruce Nuclear Power Development. The 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. Four industries with an integrated steam demand of some 1 x 10 6 lb/h were found to be prepared to consider seriously the use of nuclear steam. Their combined plants would involve a capital investment of over $200 million and provide jobs for 350-400 people. The high costs of transportation and the lack of docking facilities were considered to be the major drawbacks of the Bruce location. An indication of steam prices would be required for an over-all economic assessment

Maintenance and material aspects of DREAM reactor

International Nuclear Information System (INIS)

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

2000-01-01

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

Conceptual design of fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1984-03-01

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

Effects of cryogenic reactor irradiation on organic insulators

International Nuclear Information System (INIS)

Kato, Teruo

1995-01-01

Insulators for the superconducting magnets of fusion reactor are classified as electrical and thermal insulators for which tough organic materials will be used. When the magnet is exposed by fast neutrons and gamma-rays from plasma in a fusion reactor, the fusion reactor systems will cause fatal damage by the degradation of insulators. Therefore, it is necessary to select materials resistant irradiation damage for use as insulators. Electrical and mechanical tests were carried out at 4.2 K without warmup after the reactor irradiation at 5 K. The effects of reactor irradiation at the dose of 10 7 Gy on epoxy resins (bisphenol-A), G-10 CR, VL-E 200 and G-11 CR caused large decreases in mechanical strength. Polyetheretherketone (PEEK), polyimide and phenol novolac resins, which were used to laminate reinforced plastics with glass-cloth against irradiation, showed good resistance. Effects of cryogenic reactor irradiation on several organic materials and epoxy laminate-reinforced plastics with glass-cloth and Kevlar-cloth were also discussed. (author)

FEM analysis of mechanical behaviour of coil support connections in Wendelstein 7-X fusion reactor

International Nuclear Information System (INIS)

Krzesinski, G.; Zagrajek, T.; Marek, P.; Dobosz, R.; Czarkowski, P.; Kurzydlowski, K.J.

2006-01-01

The objective of Wendelstein 7-X project is the stellarator-type fusion reactor. In this device plasma channel is under control of magnetic field coming from magnet system of very complicated shape, made of 70 superconducting coils symmetrically arranged in 5 identical sections. Every coil is connected to central ring with two extensions which transfer loads resulting from electromagnetic field and gravity. The aim of this work was to analyse mechanical behaviour of the bolted connections using detailed 3D finite element models. All simulations were performed assuming elasto-plastic behaviour of the materials, assembly stresses and friction contacts between different parts of the connections. Stress distributions, displacements, forces acting on the bolts and welds were studied using standard and submodeling routines. The results were subsequently used to optimize the design of critical central support elements. (author)

To question of NPP power reactor choice for Kazakhstan

International Nuclear Information System (INIS)

Batyrbekov, G.A.; Makhanov, Y.M.; Reznikova, R.A.; Sidorenco, A.V.

2004-01-01

Full text: The requirements to NPP power reactors that will be under construction in Kazakhstan are proved and given in the report. A comparative analysis of the most advanced projects of power reactors with light and heavy water under pressure of large, medium and low power is carried out. Different reactors have been considered as follows: 1. Reactors with high-power (700 MW(el) and up) such as EPR, French - German reactor; CANDU-9, Canadian heavy-water reactor; System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation; APWR, Japanese advanced reactor; WWER-1000 (V-392) - development of Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor. 2. Reactors with medium power (300 MW (el) - 700 MW (el): AP-600, passive PWR of the Westinghouse company; CANDU-6, Canadian heavy-water reactor; AC-600, Chinese passive PWR; WWER-640, Russian passive reactor; MS-600 Japanese reactor of Mitsubishi Company; KSNP-600, South Korean reactor. 3. Reactors with low power (a few MW(el)- 300 MW(el)): IRIS, reactor of IV generation, developed by the International Corporation of 13 organizations from 7 countries, SMART, South Korean integrated reactor; CAREM, Argentina integrated reactor; MRX, Japanese integrated reactor; 'UNITERM', Russian NPP with integrated reactor, development of NIKIET; AHEC-80, Russian NPP, developed by OKBM. A comparison of the projects of the above-mentioned power reactors was carried out with respect to 15 criteria of nuclear, radiating, ecological safety and economic competitiveness, developed especially for this case. Data on a condition and prospects of power production and power consumption, stations and networks in Kazakhstan necessary for the choice of projects of NPP reactors for Kazakhstan are given. According to the data a balance of power production and power consumption as a whole in the country was received at the level of 59 milliard kw/h. However, strong dis balance

To question of NPP power reactor choice for Kazakhstan

International Nuclear Information System (INIS)

Batyrbekov, G.A.; Makhanov, Y.M.; Reznikova, R.A.; Sidorenco, A.V.

2004-01-01

The requirements to NPP power reactors that will be under construction in Kazakhstan are proved and given in the report. A comparative analysis of the most advanced projects of power reactors with light and heavy water under pressure of large, medium and low power is carried out. Different reactors have been considered as follows: 1. Reactors with high-power (700 MW(el) and up) such as EPR, French - German reactor; CANDU-9, Canadian heavy-water reactor; System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation; APWR, Japanese advanced reactor; WWER-1000 (V-392) - development of Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor. 2. Reactors with medium power (300 MW (el) - 700 MW (el): AP-600, passive PWR of the Westinghouse company; CANDU-6, Canadian heavy-water reactor; AC-600, Chinese passive PWR; WWER-640, Russian passive reactor; MS-600 Japanese reactor of Mitsubishi Company; KSNP-600, South Korean reactor. 3. Reactors with low power (a few MW(el)- 300 MW(el)): IRIS, reactor of IV generation, developed by the International Corporation of 13 organizations from 7 countries, SMART, South Korean integrated reactor; CAREM, Argentina integrated reactor; MRX, Japanese integrated reactor; 'UNITERM', Russian NPP with integrated reactor, development of NIKIET; AHEC-80, Russian NPP, developed by OKBM. A comparison of the projects of the above-mentioned power reactors was carried out with respect to 15 criteria of nuclear, radiating, ecological safety and economic competitiveness, developed especially for this case. Data on a condition and prospects of power production and power consumption, stations and networks in Kazakhstan necessary for the choice of projects of NPP reactors for Kazakhstan are given. According to the data a balance of power production and power consumption as a whole in the country was received at the level of 59 milliard kw/h. However, strong dis balance in the

Optimizing the use of operating experience at Ontario Hydro's Bruce Nuclear Generating Station 'A'

Energy Technology Data Exchange (ETDEWEB)

Williams, E L [Operating Experience Reactor Safety, Bruce Nuclear Generating Station ' A' , Ontario Hydro, Tiverton, Ontario (Canada)

1991-04-01

One of the most significant lessons learned from the Three Mile Island event (March 1979), and again with the Chernobyl disaster - (April 1986) was the ongoing requirement to learn from our mistakes and near misses, and those of our fellow utilities around the world: so that as an industry we do not repeat the same mistakes. The very future of our industry will depend on how well each one of us accomplishes this important ask. This paper describes in detail the challenges encountered by one station when incorporating a comprehensive 'Operating Program'. It begins with the Corporate Office's directives to its stations for such a program; and follows up with the details of the actual station implementation of the program, and day to day operating experiences. The paper describes in detail the following Operating Experience programs: - Root Cause Determination process. The Institute of Nuclear Power Operations, Human Performance Enhancement System (HPES) as an integral component of the Root Cause process. Finding solutions for our station for problems identified elsewhere is covered herein; - Significant Event Recommendation Tracking System: - Commitment Tracking System; - Operating Experience (Sharing Lessons Learned) System. The paper will show all the above processes tie closely together and complement each other. The paper discusses the staff required for such processes and their training requirements. It recommends process time lines, reporting mechanisms, and sign off requirements. It will describe the equipment utilized to carry out this work effectively, and with a minimum of staff. One unique feature of the Bruce 'A' system is an 'Effectiveness Follow-Up', usually three to six months after the event recommendations have been completed. By rechecking the finished actions and reviewing them with the personnel involved with the originating event we ensure that the real root causes have been identified and resolved. (author)

Optimizing the use of operating experience at Ontario Hydro's Bruce Nuclear Generating Station 'A'

International Nuclear Information System (INIS)

Williams, E.L.

1991-01-01

One of the most significant lessons learned from the Three Mile Island event (March 1979), and again with the Chernobyl disaster - (April 1986) was the ongoing requirement to learn from our mistakes and near misses, and those of our fellow utilities around the world: so that as an industry we do not repeat the same mistakes. The very future of our industry will depend on how well each one of us accomplishes this important ask. This paper describes in detail the challenges encountered by one station when incorporating a comprehensive 'Operating Program'. It begins with the Corporate Office's directives to its stations for such a program; and follows up with the details of the actual station implementation of the program, and day to day operating experiences. The paper describes in detail the following Operating Experience programs: - Root Cause Determination process. The Institute of Nuclear Power Operations, Human Performance Enhancement System (HPES) as an integral component of the Root Cause process. Finding solutions for our station for problems identified elsewhere is covered herein; - Significant Event Recommendation Tracking System: - Commitment Tracking System; - Operating Experience (Sharing Lessons Learned) System. The paper will show all the above processes tie closely together and complement each other. The paper discusses the staff required for such processes and their training requirements. It recommends process time lines, reporting mechanisms, and sign off requirements. It will describe the equipment utilized to carry out this work effectively, and with a minimum of staff. One unique feature of the Bruce 'A' system is an 'Effectiveness Follow-Up', usually three to six months after the event recommendations have been completed. By rechecking the finished actions and reviewing them with the personnel involved with the originating event we ensure that the real root causes have been identified and resolved. (author)

ANAEROBIC DIGESTION AND THE DENITRIFICATION IN UASB REACTOR

Directory of Open Access Journals (Sweden)

José Tavares de Sousa

2008-01-01

Full Text Available The environmental conditions in Brazil have been contributing to the development of anaerobic systems in the treatment of wastewaters, especially UASB - Upflow Anaerobic Sludge Blanket reactors. The classic biological process for removal of nutrients uses three reactors - Bardenpho System, therefore, this work intends an alternative system, where the anaerobic digestion and the denitrification happen in the same reactor reducing the number of reactors for two. The experimental system was constituted by two units: first one was a nitrification reactor with 35 L volume and 15 d of sludge age. This system was fed with raw sanitary waste. Second unit was an UASB, with 7.8 L and 6 h of hydraulic detention time, fed with ¾ of effluent nitrification reactor and ¼ of raw sanitary waste. This work had as objective to evaluate the performance of the UASB reactor. In terms of removal efficiency, of bath COD and nitrogen, it was verified that the anaerobic digestion process was not affected. The removal efficiency of organic material expressed in COD was 71%, performance already expected for a reactor of this type. It was also observed that the denitrification process happened; the removal nitrate efficiency was 90%. Therefore, the denitrification process in reactor UASB is viable.

Experience in operation of heavy water reactors

International Nuclear Information System (INIS)

Rotaru, Ion; Bilegan, Iosif; Ghitescu, Petre

1999-01-01

The paper presents the main topics of the CANDU owners group (COG) meeting held in Mangalia, Romania on 7-10 September 1998. These meetings are part of the IAEA program for exchange of information related mainly to CANDU reactor operation safety. The first meeting for PHWR reactors took place in Vienna in 1989, followed by those in Argentina (1991), India (1994) and Korea (1996). The topics discussed at the meeting in Romania were: operation experience and recent major events, performances of CANDU reactors and safe operation, nuclear safety and operation procedures of PHWR, programs and strategies of lifetime management of installations and components of NPPs, developments and updates

Medium temperature carbon dioxide gas turbine reactor

International Nuclear Information System (INIS)

Kato, Yasuyoshi; Nitawaki, Takeshi; Muto, Yasushi

2004-01-01

A carbon dioxide (CO 2 ) gas turbine reactor with a partial pre-cooling cycle attains comparable cycle efficiencies of 45.8% at medium temperature of 650 deg. C and pressure of 7 MPa with a typical helium (He) gas turbine reactor of GT-MHR (47.7%) at high temperature of 850 deg. C. This higher efficiency is ascribed to: reduced compression work around the critical point of CO 2 ; and consideration of variation in CO 2 specific heat at constant pressure, C p , with pressure and temperature into cycle configuration. Lowering temperature to 650 deg. C provides flexibility in choosing materials and eases maintenance through the lower diffusion leak rate of fission products from coated particle fuel by about two orders of magnitude. At medium temperature of 650 deg. C, less expensive corrosion resistant materials such as type 316 stainless steel are applicable and their performance in CO 2 have been proven during extensive operation in AGRs. In the previous study, the CO 2 cycle gas turbomachinery weight was estimated to be about one-fifth compared with He cycles. The proposed medium temperature CO 2 gas turbine reactor is expected to be an alternative solution to current high-temperature He gas turbine reactors

Evaluation of the passage of Lactobacillus gasseri K7 and bifidobacteria from the stomach to intestines using a single reactor model

Directory of Open Access Journals (Sweden)

von Ah Ueli

2009-05-01

Full Text Available Abstract Background Probiotic bacteria are thought to play an important role in the digestive system and therefore have to survive the passage from stomach to intestines. Recently, a novel approach to simulate the passage from stomach to intestines in a single bioreactor was developed. The advantage of this automated one reactor system was the ability to test the influence of acid, bile salts and pancreatin. Lactobacillus gasseri K7 is a strain isolated from infant faeces with properties making the strain interesting for cheese production. In this study, a single reactor system was used to evaluate the survival of L. gasseri K7 and selected bifidobacteria from our collection through the stomach-intestine passage. Results Initial screening for acid resistance in acidified culture media showed a low tolerance of Bifidobacterium dentium for this condition indicating low survival in the passage. Similar results were achieved with B. longum subsp. infantis whereas B. animalis subsp. lactis had a high survival. These initial results were confirmed in the bioreactor model of the stomach-intestine passage. B. animalis subsp. lactis had the highest survival rate (10% attaining approximately 5 × 106 cfu ml-1 compared to the other tested bifidobacteria strains which were reduced by a factor of up to 106. Lactobacillus gasseri K7 was less resistant than B. animalis subsp. lactis but survived at cell concentrations approximately 1000 times higher than other bifidobacteria. Conclusion In this study, we were able to show that L. gasseri K7 had a high survival rate in the stomach-intestine passage. By comparing the results with a previous study in piglets we could confirm the reliability of our simulation. Of the tested bifidobacteria strains, only B. animalis subsp. lactis showed acceptable survival for a successful passage in the simulation system.

Proceedings of the sixth Asian symposium on research reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-08-01

The symposium consisted of 16 sessions with 58 submitted papers. Major fields were: (1) status and future plan of research and testing reactors, (2) operating experiences, (3) design and modification of the facility, and reactor fuels, (4) irradiation studies, (5) irradiation facilities, (6) reactor characteristics and instrumentation, and (7) neutron beam utilization. Panel discussion on the 'New Trends on Application of Research and Test Reactors' was also held at the last of the symposium. About 180 people participated from China, Korea, Indonesia, Thailand, Bangladesh, Vietnam, Chinese Taipei, Belgium, France, USA, Japan and IAEA. The 58 of the presented papers are indexed individually. (J.P.N.)

Proceedings of the sixth Asian symposium on research reactors

International Nuclear Information System (INIS)

1999-08-01

The symposium consisted of 16 sessions with 58 submitted papers. Major fields were: 1) status and future plan of research and testing reactors, 2) operating experiences, 3) design and modification of the facility, and reactor fuels, 4) irradiation studies, 5) irradiation facilities, 6) reactor characteristics and instrumentation, and 7) neutron beam utilization. Panel discussion on the 'New Trends on Application of Research and Test Reactors' was also held at the last of the symposium. About 180 people participated from China, Korea, Indonesia, Thailand, Bangladesh, Vietnam, Chinese Taipei, Belgium, France, USA, Japan and IAEA. The 58 of the presented papers are indexed individually. (J.P.N.)

MIT research reactor. Power uprate and utilization

Energy Technology Data Exchange (ETDEWEB)

Hu, Lin-Wen [Nuclear Reactor Laboratory, Massachusetts Inst. of Technology, Cambridge, MA (United States)

2012-03-15

The MIT Research Reactor (MITR) is a university research reactor located on MIT campus. and has a long history in supporting research and education. Recent accomplishments include a 20% power rate to 6 MW and expanding advanced materials fuel testing program. Another important ongoing initiative is the conversion to high density low enrichment uranium (LEU) monolithic U-Mo fuel, which will consist of a new fuel element design and power increase to 7 MW. (author)

Generation IV reactors: reactor concepts

International Nuclear Information System (INIS)

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

2003-01-01

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

Protecting America Through Better Civic Education

Science.gov (United States)

2013-09-01

Studies NEA National Education Association NFTE Network for Teaching Entrepreneurship NGA National Governors Association PD#8 Presidential Directive 8...Porter, USN, who inspired me through his work, A National Strategic Narrative, with his thoughts on the intellectual capital of America, our youth—a...7 Bruce Bongar, “Defining the Need and Dexcribing the Goal,” in Psychology of Terrorism, ed. Bruce Bongar, Lisa M

Muon trackers for imaging a nuclear reactor

Science.gov (United States)

Kume, N.; Miyadera, H.; Morris, C. L.; Bacon, J.; Borozdin, K. N.; Durham, J. M.; Fuzita, K.; Guardincerri, E.; Izumi, M.; Nakayama, K.; Saltus, M.; Sugita, T.; Takakura, K.; Yoshioka, K.

2016-09-01

A detector system for assessing damage to the cores of the Fukushima Daiichi nuclear reactors by using cosmic-ray muon tomography was developed. The system consists of a pair of drift-tube tracking detectors of 7.2× 7.2-m2 area. Each muon tracker consists of 6 x-layer and 6 y-layer drift-tube detectors. Each tracker is capable of measuring muon tracks with 12 mrad angular resolutions, and is capable of operating under 50-μ Sv/h radiation environment by removing gamma induced background with a novel time-coincidence logic. An estimated resolution to observe nuclear fuel debris at Fukushima Daiichi is 0.3 m when the core is imaged from outside the reactor building.

Managing severe reactor accidents. A review and evaluation of our knowledge on reactor accidents and accident management

International Nuclear Information System (INIS)

Gustavsson, Veine

2002-11-01

The report gives a review of the results from the last years research on severe reactor accidents, and an opinion on the possibilities to refine the present strategies for accident management in Swedish and Finnish BWRs. The following aspect of reactor accidents are the major themes of the study: 1. Early pressure relief from hydrogen production; 2. Recriticality in re-flooded, degraded core; 3. Melt-through; 4. Steam explosion after melt-through; 5. Coolability of the melt after after melt-through; 6. Hydrogen fire in the reactor containment; 7. Leaking containment; 8. Hydrogen fire in the reactor building; 9. Long-time developments after a severe accident; 10. Accidents during shutdown for overhaul; 11. Information need for remedial actions. Possibilities for improving the strategies in each of these areas are discussed. The review shows that our knowledge is sufficient in the areas 1, 2, 4, 6, 8. For the other areas, more research is needed

Study of reactor parameters of on critical systems, Phase I: Safety report for RB zero power reactor; Ispitivanje reaktorskih parametara na kriticnim sistemima, I faza: Izvestaj o sigurnosti reaktora nulte snage RB

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

1962-09-15

In addition to the safety analysis for the zero power RB reactor, this report contains a general description of the reactor, reactor components, auxiliary equipment and the reactor building. Reactor Rb has been reconstructed during 1961-1962 and supplied with new safety-control system as well as with a complete dosimetry instrumentation. Since RB reactor was constructed without shielding special attention is devoted to safety and protection of the staff performing experiments. Due to changed circumstances in the Institute ( start-up of the RA 7 MW power reactor) the role of the RB reactor was redefined.

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

Science.gov (United States)

Clement, J. D.; Rust, J. H.

1977-01-01

Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

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

International Nuclear Information System (INIS)

1983-05-01

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

Licensed operating reactors: status summary report, data as of May 31, 1983. Volume 7, No. 6

International Nuclear Information System (INIS)

1983-06-01

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

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

Luncheon address: Development of the CANDU reactor

International Nuclear Information System (INIS)

Bain, A.S.

1997-01-01

The paper is a highlight of the some of the achievements in the development of the CANDU Reactor, taken from the book C anada Enters the Nuclear Age . The CANDU reactor is one of Canada's greatest scientific/engineering achievements, that started in the 1940's and bore fruit with the reactors of the 60's, 70's, and 80's. The Government decided in the 1950's to proceed with a demonstration nuclear power reactor (NPD), AECL invited 7 Canadian corporations to bid on a contract to design and construct the NPD plant. General Electric was selected. A utility was also essential for participation and Ontario Hydro was chosen. In May 1957 it was concluded that the minimum commercial size would be about 200MWe and it should use horizontal pressure tubes to contain the fuel and pressurized heavy water coolant. The book also talks of standard out-reactor components such as pumps, valves, steam generators and piping. A major in-reactor component of interest was the fuel, fuel channels and pressure tubes. A very high level of cooperation was required for the success of the CANDU program

The influence of Triga 2000 reactor operation on the surface contamination at reactor room using smear test method

International Nuclear Information System (INIS)

Bintu Khoiriyyah; Budi Purnama; Tri Cahyo Laksono

2016-01-01

The monitoring of surface contamination should be conducted to determine the safety of work areas. Surface contamination at the TRIGA 2000 reactor room which is on PSTNT-BATAN Bandung remain to be implemented although reactor not operating. In this research monitoring of surface contamination when TRIGA 2000 in operation of the first time after several years not operating aims to determine the influence on the results of monitoring. The monitoring of surface contamination has been done using smear test method at some predetermined in TRIGA 2000 reactor room. The highest surface contamination activities is obtained 0.32 Bq/cm 2 and there are some points that are not detected. Based on keputusan kepala BAPETEN No.1/Ka BAPETEN/ V/99 the work showed that the TRIGA 2000 reactor in the category of low area contamination, that is <3.7 Bq/cm 2 to gross beta. (author)

Quality assurance in the manufacture of metallic uranium fuel for research reactors

International Nuclear Information System (INIS)

Shah, B.K.; Kumar, Arbind; Nanekar, P.P.; Vaidya, P.R.

2009-01-01

Two Research Reactors viz. CIRUS and DHRUVA are operating at Trombay since 1960 and 1985 respectively. Cirus is a 40 MWth reactor using heavy water as moderator and light water as coolant. Dhruva is a 100 MWth reactor using heavy water as moderator and coolant. The maximum neutron flux of these reactors are 6.7 x 10 13 n/cm 2 /s (Cirus) and 1.8 x 10 14 n/cm 2 /s (Dhruva). Both these reactors are used for basic research, R and D in reactor technology, isotope production and operator training. Fuel material for these reactors is natural uranium metallic rods claded in finned aluminium (99.5%) tubes. This presentation will discuss various issues related to fabrication quality assurance and reactor behavior of metallic uranium fuel used in research reactors

Nuclear reactor

International Nuclear Information System (INIS)

Hattori, Sadao; Sato, Morihiko.

1994-01-01

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

The CAREM reactor and present currents in reactor design

International Nuclear Information System (INIS)

Ordonez, J.P.

1990-01-01

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

Fabrication of ORNL Fuel Irradiated in the Peach Bottom Reactor and Postirradiation Examination of Recycle Test Elements 7 and 4

International Nuclear Information System (INIS)

Long, Jr. E.L.

2001-01-01

Seven full-sized Peach Bottom Reactor fuel elements were fabricated in a cooperative effort by Oak Ridge National Laboratory (ORNL) and Gulf General Atomic (GGA) as part of the National HTGR Fuel Recycle Development Program. These elements contain bonded fuel rods and loose beds of particles made from several combinations of fertile and fissile particles of interest for present and future use in the High-Temperature Gas-Cooled Reactor (HTGR). The portion of the fuel prepared for these elements by ORNL is described in detail in this report, and it is in conjunction with the GGA report (GA-10109) a complete fabrication description of the test. In addition, this report describes the results obtained to date from postirradiation examination of the first two elements removed from the Peach Bottom Reactor, RTE-7 and -4. The fuel examined had relatively low exposure, up to about 1.5 x 10 21 neutrons/cm* fast (>0.18 MeV) fluence, compared with the peak anticipated HTGR fluence of 8.0 x 10 21 , but it has performed well at this exposure. Dimensional data indicate greater irradiation shrinkage than expected from accelerated test data to higher exposures. This suggests that either the method of extrapolation of the higher exposure data back to low exposure is faulty, or the behavior of the coated particles in the neutron spectrum characteristic of the accelerated tests does not adequately represent the behavior in an HTGR spectrum

Phosphorus removal from UASB reactor effluent by reactive media filtration.

Science.gov (United States)

Rodríguez-Gómez, Raúl; Renman, Gunno

2017-08-01

The phosphorus (P) and BOD7 removal performance of an upflow packed bed reactor (PBR) filled with two reactive filter media was studied over 50 weeks. The lower one-fifth of the reactor was filled with calcium-silicate-hydrate (Sorbulite®) and the upper four-fifth with calcium-silicate (Polonite®). A laboratory-scale upflow anaerobic sludge bed reactor (UASB) delivered wastewater to the PBR. A model was developed to describe the gradient in P concentration change in the reactor, based on reaction kinetics. The reaction terms were assumed to follow the Langmuir isotherm, based on the results obtained in a batch test. First, a comparison was made between experimental and simulated results. The capability of the model to forecast P removal capacity was then tested for three hypothetical cases: (i) reactor filled with Sorbulite and Polonite, (ii) reactor filled with only Sorbulite, and (iii) reactor filled with only Polonite. Finally, a sensitivity analysis was performed for the main parameters in the model. The average removal of P and BOD7 from the UASB effluent was 98% and 90%, respectively. The starting pH of the dual-medium effluent was 12.2 and decreased gradually over time to 11.1. The simulation both overestimated and underestimated mean measured P removal but was within the range of maximum and minimum measured values. The hypothetical cases revealed that most P was removed by Polonite due to calcium phosphate precipitation. The removal capacity of the two filter materials and their layer height in the reactor were the most sensitive parameters in the simulation.

Atomic Energy of Canada Limited annual report 2000-2001

International Nuclear Information System (INIS)

2001-01-01

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

Atomic Energy of Canada Limited annual report 2000-2001

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

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

U.S. Status of Fast Reactor Research and Technology

International Nuclear Information System (INIS)

Hill, Robert

2012-01-01

Summary: • Fast reactor R&D is focused on key technologies innovations for performance improvement (cost reduction) and safety: 1. System Integration and Concept Development; 2. Safety Technology; 3. Advanced Materials; 4. Ultrasonic Viewing; 5. Advanced Energy Conversion (Supercritical CO 2 Brayton cycle); 6. Reactor Simulation; 7. Nuclear Data; 8. Advanced Fuels. • Fast reactors have flexible capability for actinide management: – A wide variety of fuel cycle options are being considered; • International R&D collaboration pursued in Generation-IV and multilateral arrangements

Analysis of an accelerator-driven subcritical light water reactor

International Nuclear Information System (INIS)

Kruijf, W.J.M. de; Wakker, P.H.; Wetering, T.F.H. van de; Verkooijen, A.H.M.

1997-01-01

An analysis of the basic characteristics of an accelerator-driven light water reactor has been made. The waste in the nuclear fuel cycle is considerably less than in the light water reactor open fuel cycle. This is mainly caused by the use of equilibrium nuclear fuel in the reactor. The accelerator enables the use of a fuel composition with infinite multiplication factor k ∞ < 1. The main problem of the use of this type of fuel is the strongly peaked flux distribution in the reactor core. A simple analytical model shows that a large core is needed with a high peak power factor in order to generate net electric energy. The fuel in the outer regions of the reactor core is used very poorly. 7 refs., 4 figs., 1 tab

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

International Nuclear Information System (INIS)

Anon.

1986-01-01

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

Gas turbine modular helium reactor in cogeneration

International Nuclear Information System (INIS)

Leon de los Santos, G.

2009-10-01

This work carries out the thermal evaluation from the conversion of nuclear energy to electric power and process heat, through to implement an outline gas turbine modular helium reactor in cogeneration. Modeling and simulating with software Thermo flex of Thermo flow the performance parameters, based on a nuclear power plant constituted by an helium cooled reactor and helium gas turbine with three compression stages, two of inter cooling and one regeneration stage; more four heat recovery process, generating two pressure levels of overheat vapor, a pressure level of saturated vapor and one of hot water, with energetic characteristics to be able to give supply to a very wide gamma of industrial processes. Obtaining a relationship heat electricity of 0.52 and efficiency of net cogeneration of 54.28%, 70.2 MW net electric, 36.6 MW net thermal with 35% of condensed return to 30 C; for a supplied power by reactor of 196.7 MW; and with conditions in advanced gas turbine of 850 C and 7.06 Mpa, assembly in a shaft, inter cooling and heat recovery in cogeneration. (Author)

Reactor science and technology: operation and control of reactors

International Nuclear Information System (INIS)

Qiu Junlong

1994-01-01

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

Nuclear power reactors

International Nuclear Information System (INIS)

1982-11-01

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

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

Research reactors

International Nuclear Information System (INIS)

Merchie, Francois

2015-10-01

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

Fast reactors - Dounreay and the future

International Nuclear Information System (INIS)

Jordan, G.

1988-01-01

In 1960 at Dounreay, the Dounreay Fast Reactor (DFR) supplied the world's first fast reactor grid electricity, and went on to a highly successful career as a test facility, as fuel designs advanced. In the 1960s, the Prototype Fast Reactor (PFR) was designed and built, beginning operation in 1974. The PFR was built to provide a sound technical and experienced base to support the UK's future Fast Reactor development and design. The in-vessel fuel handling facilities have demonstrated the flexibility of the pool design and a considerable body of in-core fuel handling experience is available. A key issue for further Fast Reactor application is the performance of fuel and, because PFR was designed to take full-scale fuel assemblies, the fuel performance experience is directly relevant to commercial designs. The original PFR design irradiation target of 60000 MWd/t U (equivalent to 7.5 % burn-up) has already been exceeded by a factor of more than two and a 15.9 % burn-up sub-assembly has been discharged and reprocessed without difficulty. Soon a 20 % sub-assembly will follow. Also the PFR reprocessing plant has demonstrated the safety and efficiency of this essential adjunct to Fast Reactor operation. The safety and the environmental protection features of both the PFR and its fuel reprocessing plant have been demonstrated over the last 14 years. 2 refs., 3 figs

New training reactor at Dresden Technical University

International Nuclear Information System (INIS)

Hansen, W.; Knorr, J.; Wolf, T.

2006-01-01

A total of 14 low-power (up to 10 W) training reactors have been operated at German universities, 9 of them officially classified as being operational in 2004, though for very different uses. This number is expected to drop sharply. The only comprehensive upgrading of a training reactor took place at Dresden Technical University: AKR-2, the most modern facility in Germany, started routine operation in April 2005, under a newly granted license pursuant to Sec. 7, Subsec. 1 of the German Atomic Energy Act, for training students in nuclear technology, for suitable research projects, and a a center of information about reactor technology and nuclear technology for the interested public. One special aspect of this refurbishment was the installation of digital safety I and C systems of the TELEPERM XS line, which are used also in other modern plants. This fact, plus the easy possibility to use the plant for many basic experiments in reactor physics and radiation protection, make the AKR-2 attractive also to other users (e.g. for training reactor personnel or other persons working in nuclear technology). (orig.)

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

International Nuclear Information System (INIS)

Hirayama, Shoichi; Kawabe, Takaya

1995-01-01

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

Complete Sensitivity/Uncertainty Analysis of LR-0 Reactor Experiments with MSRE FLiBe Salt and Perform Comparison with Molten Salt Cooled and Molten Salt Fueled Reactor Models

Energy Technology Data Exchange (ETDEWEB)

Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mueller, Don [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-12-01

In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 ⁷LiF + BeF₂) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in ⁷Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.

Thermionic nuclear reactor systems

International Nuclear Information System (INIS)

Kennel, E.B.

1986-01-01

Thermionic nuclear reactors can be expected to be candidate space power supplies for power demands ranging from about ten kilowatts to several megawatts. The conventional ''ignited mode'' thermionic fuel element (TFE) is the basis for most reactor designs to date. Laboratory converters have been built and tested with efficiencies in the range of 7-12% for over 10,000 hours. Even longer lifetimes are projected. More advanced capabilities are potentially achievable in other modes of operation, such as the self-pulsed or unignited diode. Coupled with modest improvements in fuel and emitter material performance, the efficiency of an advanced thermionic conversion system can be extended to the 15-20% range. Advanced thermionic power systems are expected to be compatible with other advanced features such as: (1) Intrinsic subcritically under accident conditions, ensuring 100% safety upon launch abort; (2) Intrinsic low radiation levels during reactor shutdown, allowing manned servicing and/or rendezvous; (3) DC to DC power conditioning using lightweight power MOSFETS; and (4) AC output using pulsed converters

Reactor physics aspects of CANDU reactors

International Nuclear Information System (INIS)

Critoph, E.

1980-01-01

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

Research reactors

International Nuclear Information System (INIS)

Kowarski, L.

1955-01-01

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

Proceedings of the symposium on the physics and technology of reactors

International Nuclear Information System (INIS)

1993-01-01

The symposium aimed at providing the opportunity for promoting the subject and for developing the human resources in this important field in the Arab States. The symposium included 32 lectures on the following topics related to research reactors: design and development, training and operation, calculations of reactor parameters, nuclear reactions dynamics and control, reactor physics, neutron pyhsics, neutron activation analysis, in-core reactor radiation protection and shielding calculations. The lectures of the symposium were distributed over 7 sessions. An additional session was held by all participants for open discussion and recommendations

Comparison between TRU burning reactors and commercial fast reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

2001-03-01

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

Nuclear reactor types

International Nuclear Information System (INIS)

Jones, P.M.S.

1987-01-01

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

Reactor containment and reactor safety in the United States

International Nuclear Information System (INIS)

Kouts, H.

1986-01-01

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

Peptidolytic microbial community of methanogenic reactors from two modified UASBs of brewery industries

Directory of Open Access Journals (Sweden)

C. Díaz

2010-10-01

Full Text Available We studied the peptide-degrading anaerobic communities of methanogenic reactors from two mesophilic full-scale modified upflow anaerobic sludge blanket (UASB reactors treating brewery wastewater in Colombia. Most probable number (MPN counts varied between 7.1 x 10(8 and 6.6 x 10(9 bacteria/g volatile suspended solids VSS (Methanogenic Reactor 1 and 7.2 x 10(6 and 6.4 x 10(7 bacteria/g (VSS (Methanogenic Reactor 2. Metabolites detected in the highest positive MPN dilutions in both reactors were mostly acetate, propionate, isovalerate and, in some cases, negligible concentrations of butyrate. Using the highest positive dilutions of MPN counts, 50 dominant strains were isolated from both reactors, and 12 strains were selected for sequencing their 16S rRNA gene based on their phenotypic characteristics. The small-subunit rRNA gene sequences indicated that these strains were affiliated to the families Propionibacteriaceae, Clostridiaceae and Syntrophomonadaceae in the low G + C gram-positive group and Desulfovibrio spp. in the class d-Proteobacteria. The main metabolites detected in the highest positive dilutions of MPN and the presence of Syntrophomonadaceae indicate the effect of the syntrophic associations on the bioconversion of these substrates in methanogenic reactors. Additionally, the potential utilization of external electron acceptors for the complete degradation of amino acids by Clostridium strains confirms the relevance of these acceptors in the transformation of peptides and amino acids in these systems.

Fast reactor development programme in France during 1995

International Nuclear Information System (INIS)

Le Rigoleur, C.

1996-01-01

In 1995, the total amount of electricity produced in France was 471 TWh, out of which 358.2 TWh (76 %) were produced by nuclear power plants, 36.9 TWh (7.8 %) by conventional thermal plants, and 75.5 TWh (16 %) by hydraulic plants. The net electrical power consumption was 368.7 TWh. At the end of 1995, 'Electricite de France' had 54 PWR units in operation. The availability factor for these units was maintained at 81%. 1995 was marked by a decrease of unexpected shutdowns (1.8% in 1995 instead of 2.2% in 1994), a new reduction in programmed shutdown periods, and a good safety level was maintained. In the field of Fast Reactors, the main events of 1995 were the following. At the end of December 1994, the PHENIX reactor was authorized to perform its 49th cycle at 350 MW th (143 MWe). This 49th cycle was completed without any significant problems on April 7, 1995. During the remainder of the year, the reactor had been shut down in order to carry out several tasks within the scope of the ten-year extension of the PHENIX reactor's lifetime. Concerning the CREYS-MALVILLE plant (SUPER-PHENIX) the first part of the year was devoted to repairing argon leak of one of the IHX. Authorization to restart the reactor was given on August 22. The end of the year was beset by a number of minor incidents. The reactor was restarted at the end of 1995 and reactor power was increased by successive steps (30% Pn (Nominal Power) up to February 6 1996; followed by 50 %...). The 'Decret d'Autorisation de Creation' stipulates that because of its prototype character, SUPER PHENIX will have to be operated under conditions explicitly giving priority to safety and knowledge acquisition, with an objective of research and demonstration. In this context, the so-called 'knowledge acquisition' programme designed to prove the capacity of a large FBR to produce electricity on an industrial scale, to test the consumption of plutonium and minor actinides in a large fast reactor, as well as to provide

Permeated defect detecting test method and device in reactor

International Nuclear Information System (INIS)

Sakurai, Yoshishige.

1996-01-01

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

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2002-01-01

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

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2001-04-01

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

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2001-01-01

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

Reactor

International Nuclear Information System (INIS)

Toyama, Masahiro; Kasai, Shigeo.

1978-01-01

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

A stability identification system for boiling water nuclear reactors

International Nuclear Information System (INIS)

Belblidia, L.A.; Chevrier, A.

1994-01-01

Boiling water reactors are subject to instabilities under low-flow, high-power operating conditions. These instabilities are a safety concern and it is therefore important to determine stability margins. This paper describes a method to estimate a measure of stability margin, called the decay ratio, from autoregressive modelling of time series data. A phenomenological model of a boiling water reactor with known stability characteristics is used to generate time series to validate the program. The program is then applied to signals from local power range monitors from the cycle 7 stability tests at the Leibstadt plant. (author) 7 figs., 2 tabs., 12 refs

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2002-04-01

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

7Q10 flows for SRS streams

International Nuclear Information System (INIS)

Chen, K.F.

1996-01-01

The Environmental Transport Group of the Environmental Technology Section was requested to predict the seven-day ten-year low flow (7Q10 flow) for the SRS streams based on historical stream flow records. Most of the historical flow records for the SRS streams include reactor coolant water discharged from the reactors and process water released from the process facilities. The most straight forward way to estimate the stream daily natural flow is to subtract the measured upstream reactor and/or facility daily effluents from the measured downstream daily flow. Unfortunately, this method does not always work, as indicated by the fact that sometimes the measured downstream volumetric flow rates are lower than the reactor effluent volumetric flow rates. For those cases that cannot be analyzed with the simple subtracting method, an alternative method was used to estimate the stream natural flows by statistically separating reactor coolant and process water flow data. The correlation between the calculated 7Q10 flows and the watershed areas for Four Mile Branch and Pen Branch agrees with that calculated by the USGS for Upper Three Runs and Lower Three Runs Creeks. The agreement between these two independent calculations lends confidence to the 7Q10 flow calculations presented in this report

Caramel fuel for research reactors

International Nuclear Information System (INIS)

Bussy, P.

1979-11-01

This fuel for research reactors is made of UO 2 pellets in a zircaloy cladding to replace 93% enriched uranium. It is a cold fuel, non contaminating and non proliferating, enrichment is only 7 to 8%. Irradiation tests were performed until burn-up of 50000 MWD/t [fr

Coil winding pack FE-analysis for a HELIAS reactor

International Nuclear Information System (INIS)

Schauer, F.; Egorov, K.; Bykov, V.

2011-01-01

At the Max-Planck-Institut fuer Plasmaphysik (IPP) a reference design is being created of an upgraded five-periodic HELIAS type stellarator reactor which evolves from Wendelstein 7-X (W7-X) by scaling of the coil centre line geometries by a factor of four. This reactor type was extensively investigated at IPP with regard to physical characteristics and to some extent also to engineering issues. The upgrade concerns an increase of the induction at the plasma axis and correspondingly at the superconductor. The aim is to develop the reactor concept to a stage and such detail that major engineering problems are unveiled, and relevant comparisons with other concepts, including tokamaks, can be drawn in view of upcoming decisions concerning a DEMO reactor. Even though progress in plasma physics, and in particular future results of W7-X and other machines - particularly of ITER - will probably lead to somewhat different coil shapes, no principal changes of the reference design are expected. In this paper the option of a roll-formed square coil cable jacket is investigated. Detailed structural FE analysis of the coil winding pack demonstrates the feasibility of such a conductor which appears to be the most economical option. It also allows sufficient space for a cable current density very similar to that of the ITER TF coil with a similar overall winding pack cross section of ∼0.5 m 2 . Already existing Nb 3 Sn conductors could thus be safely applied in such a HELIAS reactor. Obvious progress of superconductor technology, particularly concerning Nb 3 Al, will be beneficial concerning savings of conductor material, ease of manufacture, higher operation temperature, etc.

Research reactor standards and their impact on the TRIGA reactor community

International Nuclear Information System (INIS)

Richards, W.J.

1980-01-01

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

Results of the Nucifer reactor neutrino experiment

Energy Technology Data Exchange (ETDEWEB)

Buck, Christian; Lindner, Manfred [MPIK Heidelberg (Germany)

2016-07-01

Nuclear reactors are a strong and pure source of electron antineutrinos. With neutrino experiments close to compact reactor cores new insights into neutrino properties and reactor physics can be obtained. The Nucifer experiment is one of the pioneers in this class of very short baseline projects. Its detector to reactor distance is only about 7 m. The data obtained in the last years allowed to estimate the plutonium concentration in the reactor core by the neutrino flux measurement. This is of interest for safeguard applications and non proliferation efforts. The antineutrinos in Nucifer are detected via the inverse beta decay on free protons. Those Hydrogen nuclei are provided by 850 liters of organic liquid scintillator. For higher detection efficiency and background reduction the liquid is loaded with Gadolinium. Despite all shielding efforts and veto systems the background induced by the reactor activity and cosmogenic particles is still the main challenge in the experiment. The principle of the Nucifer detector is similar to the needs of upcoming experiments searching for sterile neutrinos. Therefore, the Nucifer results are also valuable input for the understanding and optimization of those next generation projects. The observation of sterile neutrinos would imply new physics beyond the standard model.

Statement to International Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities, 7 December 2009, Kyoto, Japan

International Nuclear Information System (INIS)

Amano, Yukiya

2009-01-01

Full text: Distinguished Guests, Ladies and Gentlemen, It is my honour to address participants at this opening session of the International Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities, organized by the IAEA and hosted by the Japan Atomic Energy Agency. Fast reactor technology has the potential to ensure that energy resources which would last hundreds of years with the technology we are using today will actually last several thousand years. In other words it can withstand enormous increases in demand. This innovative technology also reduces the risk to the environment and helps to limit the burden that will be placed on future generations in the form of waste products. The coming year will be an exciting one for the development of fast-spectrum nuclear reactors. We expect to reach many important milestones: - the first criticality of the China Experimental Fast Reactor; - the restart of the Monju prototype fast reactor in Japan; and - the new insights we will gain through the end-of-life studies at the Phenix reactor in France. In the near future, new fast reactors will be commissioned: the 500MW(e) Prototype Fast Breeder Reactor in India, the first in a series of five of the same type, and the BN-800 reactor in the Russian Federation. Moreover, France, Japan, India, China and the Republic of Korea are preparing advanced prototypes, demonstration or commercial reactors for the 2020-2030 period. Nuclear power is set to be an increasingly important part of the global energy mix in the coming decades as demand for energy grows. Scores of countries in both the developed and developing world have told the IAEA that they are interested in introducing nuclear power. The 30 countries which already have nuclear power reactors are set to build more. This trend is likely to be accompanied by accelerated deployment of fast reactors. Continued advances in research and technology development are necessary to ensure improved economics and

Reactor operation

CERN Document Server

Shaw, J

2013-01-01

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

Improved nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding liquid metal coolant and housing the core within the pool. A generally cylindrical concrete containment structure surrounds the reactor vessel and a central support pedestal is anchored to the containment structure base mat and supports the bottom wall of the reactor vessel and the reactor core. The periphery of the reactor vessel bore is supported by an annular structure which allows thermal expansion but not seismic motion of the vessel, and a bed of thermally insulating material uniformly supports the vessel base whilst allowing expansion thereof. A guard ring prevents lateral seismic motion of the upper end of the reactor vessel. The periphery of the core is supported by an annular structure supported by the vessel base and keyed to the vessel wall so as to be able to expand but not undergo seismic motion. A deck is supported on the containment structure above the reactor vessel open top by annular bellows, the deck carrying the reactor control rods such that heating of the reactor vessel results in upward expansion against the control rods. (author)

Reactor as furnace and reactor as lamp

International Nuclear Information System (INIS)

Goldanskii, V.I.

1992-01-01

There are presented general characteristics of the following ways of transforming of nuclear energy released in reactors into chemical : ordinary way (i.e. trough the heat, mechanical energy and electricity); chemonuclear synthesis ; use of high-temperature fuel elements (reactor as furnace); use of the mixed nγ-radiation of reactors; use of the radiation loops; radiation - photochemical synthesis (reactor as lamp). Advantage and disadvantages of all above variants are compared. The yield of the primary product of fixation of nitrogen (nitric oxide NO) in reactor with the high-temperature (above ca. 1900degC) fuel elements (reactor-furnace) can exceed W ∼ 200 kg per gram of burned uranium. For the latter variant (reactor-lamp) the yield of chemical products can reach W ∼ 60 kg. per gram of uranium. Such values of W are close to or even strongly exceed the yields of chemical products for other abovementioned variants and - what is particularly important - are not connected to the necessity of archscrupulous removal of radioactive contamination of products. (author)

Development of Reactor Console Simulator for PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Mohd Idris Taib; Izhar Abu Hussin; Mohd Khairulezwan Abdul Manan; Nufarhana Ayuni Joha; Mohd Sabri Minhat

2012-01-01

The Reactor Console Simulator will be an interactive tool for operator training and teaching of PUSPATI TRIGA Reactor. Behaviour and characteristic for reactor console and reactor itself can be evaluated and understand. This Simulator will be used as complement for actual present reactor console. Implementation of man-machine interface is using computer screens, keyboard and mouse. Multiple screens are used to match the physical of present reactor console. LabVIEW software are using for user interface and mathematical calculation. Polynomial equation based on control rods calibration data as well as operation parameters record was used to calculate the estimated reactor console parameters. (author)

Tritium release during inspection of reactor 'RA' at 'Vinca' institute

International Nuclear Information System (INIS)

Sipka, V.; Miljevic, N.; Grsic, Z.; Todorovic, D.; Radenkovic, M.

1997-01-01

Tritium content in daily precipitation, atmospheric water vapor inside of the reactor hall and around 'Vinca' Institute as well as in soil up to 800 m distance was monitoring during the regular inspection of the fuel channels. Tritium activity in the reactor hall air moisture was in the range from 0.022 to 6.7 MBq/m 3 . Tritium content in soil moisture between 12.7 and 530.9 Bq/l indicate a certain contamination due to tritium release in the environment, depending on the depth and distance from the place of release (author) [sr

Reactor safeguards

CERN Document Server

Russell, Charles R

1962-01-01

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

A design study of reactor core optimization for direct nuclear heat-to-electricity conversion in a space power reactor

Energy Technology Data Exchange (ETDEWEB)

Yoshikawa, Hidekazu; Takahashi, Makoto; Shimoda, Hiroshi; Takeoka, Satoshi [Kyoto Univ. (Japan); Nakagawa, Masayuki; Kugo, Teruhiko

1998-01-01

To propose a new design concept of a nuclear reactor used in the space, research has been conducted on the conceptual design of a new nuclear reactor on the basis of the following three main concepts: (1) Thermionic generation by thermionic fuel elements (TFE), (2) reactivity control by rotary reflector, and (3) reactor cooling by liquid metal. The outcomes of the research are: (1) A calculation algorithm was derived for obtaining convergent conditions by repeating nuclear characteristic calculation and thermal flow characteristic calculation for the space nuclear reactor. (2) Use of this algorithm and the parametric study established that a space nuclear reactor using 97% enriched uranium nitride as the fuel and lithium as the coolant and having a core with a radius of about 25 cm, a height of about 50 cm and a generation efficiency of about 7% can probably be operated continuously for at least more than ten years at 100 kW only by reactivity control by rotary reflector. (3) A new CAD/CAE system was developed to assist design work to optimize the core characteristics of the space nuclear reactor comprehensively. It is composed of the integrated design support system VINDS using virtual reality and the distributed system WINDS to collaboratively support design work using Internet. (N.H.)

Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

International Nuclear Information System (INIS)

Greneche, D.

2014-01-01

This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

Recent fuel handling experience in Canada

International Nuclear Information System (INIS)

Welch, A.C.

1991-01-01

For many years, good operation of the fuel handling system at Ontario Hydro's nuclear stations has been taken for granted with the unavailability of the station arising from fuel handling system-related problems usually contributing less than one percent of the total unavailability of the stations. While the situation at the newer Hydro stations continues generally to be good (with the specific exception of some units at Pickering B) some specific and some general problems have caused significant loss of availability at the older plants (Pickering A and Bruce A). Generally the experience at the 600 MWe units in Canada has also continued to be good with Point Lepreau leading the world in availability. As a result of working to correct identified deficiencies, there were some changes for the better as some items of equipment that were a chronic source of trouble were replaced with improved components. In addition, the fuel handling system has been used three times as a delivery system for large-scale non destructive examination of the pressure tubes, twice at Bruce and once at Pickering and performing these inspections this way has saved many days of reactor downtime. Under COG there are several programs to develop improved versions of some of the main assemblies of the fuelling machine head. This paper will generally cover the events relating to Pickering in more detail but will describe the problems with the Bruce Fuelling Machine Bridges since the 600 MW 1P stations have a bridge drive arrangement that is somewhat similar to Bruce

Control of reactor coolant flow path during reactor decay heat removal

International Nuclear Information System (INIS)

Hunsbedt, A.N.

1988-01-01

This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

Nuclear reactors for space electric power

International Nuclear Information System (INIS)

Buden, D.

1978-06-01

The Los Alamos Scientific Laboratory is studying reactor power plants for space applications in the late 1980s and 1990s. The study is concentrating on high-temperature, compact, fast reactors that can be coupled with various radiation shielding systems and thermoelectric, dynamic, or thermionic electric power conversion systems, depending on the mission. Lifetimes of 7 to 10 yr at full power, at converter operating temperatures of 1275 to 1675 0 K, are being studied. The systems are being designed such that no single-failure modes exist that will cause a complete loss of power. In fact, to meet the long lifetimes, highly redundant design features are being emphasized. Questions have been raised about safety since the COSMOS 954 incident. ''Fail-safe'' means to prevent exposure of the population to radioactive material, meeting the environmental guidelines established by the U.S. Government have been and continue to be a necessary requirement for any space reactor program. The major safety feature to prevent prelaunch and launch radioactive material hazards is not operating the reactor before achieving the prescribed orbit. Design features in the reactor ensure that accidental criticality cannot occur. High orbits (above 400 to 500 nautical miles) have sufficient lifetimes to allow radioactive elements to decay to safe levels. The major proposed applications for satellites with reactors in Earth orbit are in geosynchronous orbit (19,400 nautical miles). In missions at geosynchronous orbit, where orbital lifetimes are practically indefinite, the safety considerations are negligible. Orbits below 400 to 500 nautical miles are the ones where a safety issue is involved in case of satellite malfunction. The potential missions, the question of why reactors are being considered as a prime power candidate, reactor features, and safety considerations will be discussed

Material and component progress within ARCHER for advanced high temperature reactor

International Nuclear Information System (INIS)

Buckthorpe, D.E.; Davies, M.; Pra, F.; Bonnamy, P.; Fokkens, J.; Heijna, M.; Bout, N. de; Vreeling, A.; Bourlier, F.; Lhachemi, D.; Woayehune, A.; Dubiez-le-Goff, S.; Hahner, P.; Futterer, M.; Berka, J.; Kalivodora, J.; Pouchon, M.A.; Schmitt, R.; Homerin, P.; Marsden, B.; Mummery, P.; Mutch, G.; Ponca, D.; Buhl, P.; Hoffmann, M.; Rondet, F.; Pecherty, A.; Baurand, F.; Alenda, F.; Esch, M.; Kohlz, N.; Reed, J.; Fachinger, J.; Klower, Dr.

2014-01-01

The ARCHER (Advanced High-Temperature Reactors for Cogeneration of Heat and Electricity R and D) integrated project started in 2011 as part of the European Commission 7. Framework Programme (FP7) for a period of four years to perform High Temperature Reactor technology R and D in support of reactor demonstration. The project consortium encompasses conventional and Nuclear Industry, Utilities, Technical Support Organizations, Research and Development Organizations and Academia. The activities involved contribute to the Generation IV (GIF) International Forum and collaborate with related projects in the US, China, Japan, and the Republic of Korea in cooperation with IAEA and ISTC. This paper addresses the progress of the work on materials and component technologies within ARCHER over the first two years of the project. (authors)

RB research nuclear reactor, Annual report for 1982; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1981. godini

Energy Technology Data Exchange (ETDEWEB)

Markovic, H; Pesic, M; Vranic, S; Petronijevic, M; Zivkovic, B [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

1982-12-15

This report includes data concerned with reactor operation and utilization, status of reactor components and equipment, refurbishment of the equipment, dosimetry and radiation protection, reactor staff, financing. It includes 7 Annexes as follows: Maintenance of reactor equipment in 1982; contents of the RB reactor safety report; review of radiation doses in the reactor building and exposure doses for the reactor staff; utilization of the RB reactor in 1982; and financial data.

The research reactors their contribution to the reactors physics

International Nuclear Information System (INIS)

Barral, J.C.; Zaetta, A.; Johner, J.; Mathoniere, G.

2000-01-01

The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

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)

Handling of spent fuel from research reactors in Japan

International Nuclear Information System (INIS)

Kanda, K.

1997-01-01

In Japan eleven research reactors are in operation. After the 19th International Meeting on Reduced Enrichment for Research Reactors and Test Reactors (RERTR) on October 6-10, 1996, Seoul, Korea, the Five Agency Committee on Highly Enriched Uranium, which consists of Science and Technology Agency, the Ministry of Education, Science and Culture, the Ministry of Foreign Affairs, Japan Atomic Energy Research Institute (JAERI) and Kyoto University Research Reactor Institute (KURRI) met on November 7,1996, to discuss the handling of spent fuel from research reactors in Japan. Advantages and disadvantages to return spent fuel to the USA in comparison to Europe were discussed. So far, a number of spent fuel elements in JAERI and KURRI are to be returned to the US. The first shipment to the US is planned for 60 HEU elements from JMTR in 1997. The shipment from KURRI is planned to start in 1999. (author)

A neutronics feasibility study for the LEU conversion of Poland's Maria research reactor

International Nuclear Information System (INIS)

Bretscher, M. M.

1998-01-01

The MARIA reactor is a high-flux multipurpose research reactor which is water-cooled and moderated with both beryllium and water. Standard HEU (80% 235 U)fuel assemblies consist of six concentric fuel tubes of a U-Al alloy clad in aluminum. Although the inventory of HEU (80%) fuel is nearly exhausted, a supply of highly-loaded 36%-enriched fuel assemblies is available at the reactor site. Neutronic equilibrium studies have been made to determine the relative performance of fuels with enrichments of 80%, 36% and 19.7%. These studies indicate that LEU (19.7%) densities of about 2.5 gU/cm 3 and 3.8 gU/cm 3 are required to match the performance of the MARIA reactor with 80%-enriched and with 36%-enriched fuels, respectively

Reactor physics and reactor computations

International Nuclear Information System (INIS)

Ronen, Y.; Elias, E.

1994-01-01

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

Mitigate Strategy of Very High Temperature Reactor Air-ingress Accident

Energy Technology Data Exchange (ETDEWEB)

Ham, Tae Kyu [KHNP CRI, Daejeon (Korea, Republic of); Arcilesi, David J.; Sun, Xiaodong; Christensen, Richard N. [The Ohio State University, Columbus (United States); Oh, Chang H.; Kim, Eung S. [Idaho National Laboratory, Idaho (United States)

2016-10-15

A critical safety event of the Very High Temperature Reactor (VHTR) is a loss-of-coolant accident (LOCA). Since a VHTR uses graphite as a core structure, if there is a break on the pressure vessel, the air in the reactor cavity could ingress into the reactor core. The worst case scenario of the accident is initiated by a double-ended guillotine break of the cross vessel that connects the reactor vessel and the power conversion unit. The operating pressures in the vessel and containment are about 7 and 0.1 MPa, respectively. In the VHTR, the reactor pressure vessel is located within a reactor cavity which is filled with air during normal operation. Therefore, the air-helium mixture in the cavity may ingress into the reactor pressure vessel after the depressurization process. In this paper, a commercial computational fluid dynamics (CFD) tool, FLUENT, was used to figure out air-ingress mitigation strategies in the gas-turbine modular helium reactor (GT-MHR) designed by General Atomics, Inc. After depressurization, there is almost no air in the reactor cavity; however, the air could flow back to the reactor cavity since the reactor cavity is placed in the lowest place in the reactor building. The heavier air could flow to the reactor cavity through free surface areas in the reactor building. Therefore, Argon gas injection in the reactor cavity is introduced. The injected argon would prevent the flow by pressurizing the reactor cavity initially, and eventually it prevents the flow by making the gas a heavier density than air in the reactor cavity. The gate opens when the reactor cavity is pressurized during the depressurization and it closes by gravity when the depressurization is terminated so that it can slow down the air flow to the reactor cavity.

Spent fuel bundle counter sequence error manual - BRUCE NGS

International Nuclear Information System (INIS)

Nicholson, L.E.

1992-01-01

The Spent Fuel Bundle Counter (SFBC) is used to count the number and type of spent fuel transfers that occur into or out of controlled areas at CANDU reactor sites. However if the transfers are executed in a non-standard manner or the SFBC is malfunctioning, the transfers are recorded as sequence errors. Each sequence error message typically contains adequate information to determine the cause of the message. This manual provides a guide to interpret the various sequence error messages that can occur and suggests probable cause or causes of the sequence errors. Each likely sequence error is presented on a 'card' in Appendix A. Note that it would be impractical to generate a sequence error card file with entries for all possible combinations of faults. Therefore the card file contains sequences with only one fault at a time. Some exceptions have been included however where experience has indicated that several faults can occur simultaneously

Reactor water spontaneous circulation structure in reactor pressure vessel

International Nuclear Information System (INIS)

Takahashi, Kazumi

1998-01-01

The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

The heavy water accountancy for research reactors in JAERI

International Nuclear Information System (INIS)

Yoshijima, Tetsuo; Tanaka, Sumitoshi; Nemoto, Denjirou

1998-11-01

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

Reactor physics tests of TRIGA Mark-II Reactor in Ljubljana

International Nuclear Information System (INIS)

Ravnik, M.; Mele, I.; Trkov, A.; Rant, J.; Glumac, B.; Dimic, V.

2008-01-01

TRIGA Mark-II Reactor in Ljubljana was recently reconstructed. The reconstruction consisted mainly of replacing the grid plates, the control rod mechanisms and the control unit. The standard type control rods were replaced by the fuelled follower type, the central grid location (A ring) was adapted for fuel element insertion, the triangular cutouts were introduced in the upper plate design. However, the main novelty in reactor physics and operational features of the reactor was the installation of a pulse rod. Having no previous operational experience in pulsing, a detailed and systematic sequence of tests was defined in order to check the predicted design parameters of the reactor with measurements. The following experiments are treated in this paper: initial criticality, excess reactivity measurements, control rod worth measurement, fuel temperature distribution, fuel temperature reactivity coefficient, pulse parameters measurement (peak power, prompt energy, peak temperature). Flux distributions in steady state and pulse mode were measured as well, however, they are treated only briefly due to the volume of the results. The experiments were performed with completely fresh fuel of 12 w% enriched Standard Stainless Steel type. The core configuration was uniform (one fuel element type, including fuelled followers) and compact (no irradiation channels or gaps), as such being particularly convenient for testing the computer codes for TRIGA reactor calculations. Comparison of analytical predictions, obtained with WIMS, SLXTUS, TRIGAP and PULSTRI codes to measured values showed agreement within the error of the measurement and calculation. The paper has the following contents: 1. Introduction; 2. Steady State Experiments; 2.1. Core loading and critical experiment; 2.2. Flux range determination for tests at zero power; 2.3. Digital reactivity meter checkout; 2.4. Control rod worth measurements; 2.5. Excess reactivity measurement; 2.6. Thermal power calibration; 2.7

United States Domestic Research Reactor Infrastructure - TRIGA Reactor Fuel Support

International Nuclear Information System (INIS)

Morrell, Douglas

2008-01-01

The purpose of the United State Domestic Research Reactor Infrastructure Program is to provide fresh nuclear reactor fuel to United States universities at no, or low, cost to the university. The title of the fuel remains with the United States government and when universities are finished with the fuel, the fuel is returned to the United States government. The program is funded by the United States Department of Energy - Nuclear Energy division, managed by Department of Energy - Idaho Field Office, and contracted to the Idaho National Laboratory's Management and Operations Contractor - Battelle Energy Alliance. Program has been at Idaho since 1977 and INL subcontracts with 26 United States domestic reactor facilities (13 TRIGA facilities, 9 plate fuel facilities, 2 AGN facilities, 1 Pulstar fuel facility, 1 Critical facility). University has not shipped fuel since 1968 and as such, we have no present procedures for shipping spent fuel. In addition: floor loading rate is unknown, many interferences must be removed to allow direct access to the reactor tank, floor space in the reactor cell is very limited, pavement ends inside our fence; some of the surface is not finished. The whole approach is narrow, curving and downhill. A truck large enough to transport the cask cannot pull into the lot and then back out (nearly impossible / refused by drivers); a large capacity (100 ton), long boom crane would have to be used due to loading dock obstructions. Access to the entrance door is on a sidewalk. The campus uses it as a road for construction equipment, deliveries and security response. Large trees are on both sides of sidewalk. Spent fuel shipments have never been done, no procedures approved or in place, no approved casks, no accident or safety analysis for spent fuel loading. Any cask assembly used in this facility will have to be removed from one crane, moved on the floor and then attached to another crane to get from the staging area to the reactor room. Reactor

Requirements for light water reactors

International Nuclear Information System (INIS)

Hedin, F.

2009-01-01

The EUR (European Utilities Requirements) is an organization founded in 1991 whose aim was to write down the European specifications and requirements for the future reactors of third generation. EUR gathers most of the nuclear power producers of Europe. The EUR document has been built on the large and varied experience of EUR members and can be used to elaborate invitations to tender for nuclear projects. 4000 requirements only for the nuclear part of the plant are listed, among which we have: -) the probability of core meltdown for a reactor must be less than 10 -6 per year, -) the service life of every component that is not replaceable must be 60 years, -) the capacity of the spent fuel pool must be sufficient to store 10-15 years of production without clearing out. The EUR document is both open and complete: every topic has been considered, it does not favor any type of reactor but can ban any technology that is too risky or has an unfavourable feedback experience. The assessment of the conformity with the EUR document of 7 reactor projects (BWR 90/, EPR, EP1000, SWR1000, ABWR, AP1000 and VVER-AES-92) has already be made. (A.C.)

Reactor

International Nuclear Information System (INIS)

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

1979-01-01

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

Reactor Physics Training

International Nuclear Information System (INIS)

Baeten, P.

2007-01-01

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

FBR type reactor

International Nuclear Information System (INIS)

Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

1994-01-01

In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

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 container

International Nuclear Information System (INIS)

Kato, Masami; Nishio, Masahide.

1987-01-01

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

Review of the United Kingdom fast reactor programme - March 1986

International Nuclear Information System (INIS)

Bramman, J.I.; John, C.T.; Wheeler, R.C.

1986-01-01

The UK programme in the field of fast reactors has continued successfully towards the following main objectives, details of which are contained in subsequent sections of this report: (2) progress with the prototype fast reactor (PFR) which achieved its design power on 4 March 1985; (3) nuclear fuel reprocessing; (4) commercial design studies; (5) structural integrity of LMFBR during its lifetime; (6) R and D work on components of LMFBR; (7) materials study; (8) sodium chemistry; (9) reactor core and fuel design philosophy; (10) safety problems; (11) plant performance studies

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

The Integration Of Process Heat Applications To High Temperature Gas Reactors

International Nuclear Information System (INIS)

McKellar, Michael G.

2011-01-01

A high temperature gas reactor, HTGR, can produce industrial process steam, high-temperature heat-transfer gases, and/or electricity. In conventional industrial processes, these products are generated by the combustion of fossil fuels such as coal and natural gas, resulting in significant emissions of greenhouse gases such as carbon dioxide. Heat or electricity produced in an HTGR could be used to supply process heat or electricity to conventional processes without generating any greenhouse gases. Process heat from a reactor needs to be transported by a gas to the industrial process. Two such gases were considered in this study: helium and steam. For this analysis, it was assumed that steam was delivered at 17 MPa and 540 C and helium was delivered at 7 MPa and at a variety of temperatures. The temperature of the gas returning from the industrial process and going to the HTGR must be within certain temperature ranges to maintain the correct reactor inlet temperature for a particular reactor outlet temperature. The returning gas may be below the reactor inlet temperature, ROT, but not above. The optimal return temperature produces the maximum process heat gas flow rate. For steam, the delivered pressure sets an optimal reactor outlet temperature based on the condensation temperature of the steam. ROTs greater than 769.7 C produce no additional advantage for the production of steam.

Socioeconomic consequences of nuclear reactor accidents

International Nuclear Information System (INIS)

Tawil, J.J.; Callaway, J.W.; Coles, B.L.; Cronin, F.J.; Currie, J.W.; Imhoff, K.L.; Lewis, P.M.; Nesse, R.J.; Strenge, D.L.

1984-06-01

This report identifies and characterizes the off-site socioeconomic consequences that would likely result from a severe radiological accident at a nuclear power plant. The types of impacts that are addressed include economic impacts, health impacts, social/psychological impacts and institutional impacts. These impacts are identified for each of several phases of a reactor accident - from the warning phase through the post-resettlement phase. The relative importance of the impact during each accident phase and the degree to which the impact can be predicted are indicated. The report also examines the methods that are currently used for assessing nuclear reactor accidents, including development of accident scenarios and the estimating of socioeconomic accident consequences with various models. Finally, a critical evaluation is made regarding the use of impact analyses in estimating the contribution of socioeconomic consequences to nuclear accident reactor accident risk. 116 references, 7 figures, 15 tables

Low power reactor for remote applications

International Nuclear Information System (INIS)

Meier, K.L.; Palmer, R.G.; Kirchner, W.L.

1985-01-01

A compact, low power reactor is being designed to provide electric power for remote, unattended applications. Because of the high fuel and maintenance costs for conventional power sources such as diesel generators, a reactor power supply appears especially attractive for remote and inaccessible locations. Operating at a thermal power level of 135 kWt, the power supply achieves a gross electrical output of 25 kWe from an organic Rankine cycle (ORC) engine. By intentional selection of design features stressing inherent safety, operation in an unattended mode is possible with minimal risk to the environment. Reliability is achieved through the use of components representing existing, proven technology. Low enrichment uranium particle fuel, in graphite core blocks, cooled by heat pipes coupled to an ORC converter insures long-term, virtually maintenance free, operation of this reactor for remote applications. 10 refs., 7 figs., 3 tabs

The University of Toronto's lasting contribution to war surgery: how Maj. L. Bruce Robertson fundamentally transformed thinking toward blood transfusion during the First World War.

Science.gov (United States)

Tien, Abigail; Beckett, Andrew; Pannell, Dylan

2017-06-01

During the Great War, Canadian military surgeons produced some of the greatest innovations to improve survival on the battlefield. Arguably, the most important was bringing blood transfusion practice close to the edge of the battlefield to resuscitate the many casualties dying of hemorrhagic shock. Dr. L. Bruce Robertson of the Canadian Army Medical Corps was the pioneering surgeon from the University of Toronto who was able to demonstrate the benefit of blood transfusions near the front line and counter the belief that saline was the resuscitation fluid of choice in military medicine. Robertson would go on to survive the Great War, but would be taken early in life by influenza. Despite his life and career being cut short, Robertson's work is still carried on today by many military medical organizations who strive to bring blood to the wounded in austere and dangerous settings. This article has an Appendix, available at canjsurg.ca.

Water Chemistry Control in Reducing Corrosion and Radiation Exposure at PWR Reactor

International Nuclear Information System (INIS)

Febrianto

2006-01-01

Water chemistry control plays an important role in relation to plant availability, reliability and occupational radiation exposures. Radiation exposures of nuclear plant workers are determined by the radiation rate dose and by the amount maintenance and repair work time Water chemistry has always been, from beginning of operation of power Pressurized Water Reactor, an important factor in determining the integrity of reactor components, fuel cladding integrity and minimize out of core radiation exposures. For primary system, the parameters to control the quality of water chemistry have been subject to change in time. Reactor water coolant pH need to be optimally controlled and be operated in range pH 6.9 to 7.4. At pH lower than 6.9, cause increasing the radiation exposure level and increasing coolant water pH higher than 7.4 will decrease radiation exposure level but increasing risk to fuel cladding and steam generator tube. Since beginning 90 decade, PWR water coolant pH tend to be operated at pH 7.4. This paper will discuss concerning water chemistry development in reducing corrosion and radiation exposure dose in PWR reactor. (author)

Office for Analysis and Evaluation of Operational Data 1992 annual report: Power reactors. Volume 7, No. 1

Energy Technology Data Exchange (ETDEWEB)

None

1993-07-01

The annual report of the US Nuclear Regulatory Commission`s Office for Analysis and Evaluation of Operational Data (AEOD) is devoted to the activities performed during 1992. The report is published in two separate parts. NUREG-1272, Vol. 7, No. 1, covers power reactors and presents an overview of the operating experience of the nuclear power industry from the NRC perspective, including comments about the trends of some key performance, measures. The report also includes the principal findings and issues identified in AEOD studies over the past year, and summarizes information from such sources as licensee event report% diagnostic evaluations, and reports to the NRC`s Operations Center. The reports contain a discussion of the Incident Investigation Team program and summarize the Incident Investigation Team and Augmented Inspection Team reports for that group of licensees. NUREG-1272, Vol. 7, No. 2, covers nonreactors and presents a review of the events and concerns during 1992 associated with the use of licensed material in nonreactor applications, such as personnel overexposures and medical misadministrations. Each volume contains a list of the AEOD reports issued for 1984--1992.

Nuclear Reactor RA Safety Report, Format and Contents

International Nuclear Information System (INIS)

1986-11-01

This is a new complete version of the safety report of nuclear reactor RA is made according to the recommendations of the IAEA. Report includes all the relevant data needed for evaluation of safe operation of this nuclear facility. Each of seven volumes of this report cover separate topics as follows: (1) introduction; (2) Site characteristics; (3) description of the reactor building and installations; (4) description of the reactor; (5) description of the coolant system; (6) description of the regulation and safety instrumentation; (7) description of the power supply system; (8) description of the auxiliary systems; (9) radiation protection issues; (10) radioactive waste management (11) reactor operation; (12) accident analysis during previous operation; (13) analysis of possible accident causes; (14) safety analysis and preventive actions: (15) analysis of significant accidents; (16) analysis of maximum possible accident; (17) environmental impact analysis in case of accident [sr

Simulation of the Point Lepreau core-follow history with SORO

Energy Technology Data Exchange (ETDEWEB)

Shanes, F.C.; Olive, C.G.; Cheng, I. [Nuclear Safety Solutions Limited, Toronto, Ontario (Canada); Banica, C. [Ontario Power Generation, Ajax, Ontario (Canada); Newman, C. [NB Power Nuclear, Point Lepreau Generating Station, Lepreau, New Brunswick (Canada); Nainer, O. [Bruce Power, Toronto, Ontario (Canada)

2006-07-01

The core tracking SORO program calculates the flux, burn-ups, bundle and channel powers, and is used for Ontario Power Generation and Bruce Power reactors. This paper presents the results of a comparison of the SORO cell fluxes against the Point Lepreau CANDU-6 measured flux data. A SORO model was created for the Point Lepreau reactor, and simulations were carried out to compare the SORO fluxes to the Travelling Flux Detector (TFD) scans and one year of operating history. Considering the good agreement between measured and computed fluxes, the results provide confidence that SORO accurately calculates the cell fluxes and bundle powers. (author)

Simulation of the Point Lepreau core-follow history with SORO

International Nuclear Information System (INIS)

Shanes, F.C.; Olive, C.G.; Cheng, I.; Banica, C.; Newman, C.; Nainer, O.

2006-01-01

The core tracking SORO program calculates the flux, burn-ups, bundle and channel powers, and is used for Ontario Power Generation and Bruce Power reactors. This paper presents the results of a comparison of the SORO cell fluxes against the Point Lepreau CANDU-6 measured flux data. A SORO model was created for the Point Lepreau reactor, and simulations were carried out to compare the SORO fluxes to the Travelling Flux Detector (TFD) scans and one year of operating history. Considering the good agreement between measured and computed fluxes, the results provide confidence that SORO accurately calculates the cell fluxes and bundle powers. (author)

The rehabilitation/upgrading of Philippine Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Renato, T Banaga [Philippines Nuclear Research Inst., Quezon (Philippines)

1998-10-01

The Philippine Research Reactor (PRR-1) is the only research reactor in the Philippines. It was acquired through the Bilateral Agreement with the United States of America. The General Electric (G.E.) supplied PRR-1 first become operational in 1963 and used MTR plate type fuel. The original one-megawatt G.E. reactor was shutdown and converted into a 3 MW TRIGA PULSING REACTOR in 1984. The conversion includes the upgrading of the cooling system, replacement of new reactor coolant pumps, heat exchanger, cooling tower, replacement of new nuclear instrumentation and standard TRIGA console, TRIGA fuel supplied by General Atomic (G.A.). Philippine Nuclear Research Institute (PNRI) provided the old reactor, did the detailed design of the new cooling system, provided the new non-nuclear instrumentation and electrical power supply system and performed all construction, installation and modification work on site. The TRIGA conversion fuel is contained in a shrouded 4-rod cluster which fit into the original grid plate. The new fuel is a E{sub 1}-U-Z{sub 1}-H{sub 1.6} TRIGA fuel, has a 20% wt Uranium loading with 19.7% U-235 enrichment and about 0.5 wt % Erbium. The Start-up, calibration and Demonstration of Pulsing and Full Power Operation were completed during a three week start-up phase which were performed last March 1968. A few days after, a leak in the pool liner was discovered. The reactor was shutdown again for repair and up to present the reactor is still in the process of rehabilitation. This paper will describe the rehabilitation/upgrading done on the PRR-1 since 1988 up to present. (author)

The rehabilitation/upgrading of Philippine Research Reactor

International Nuclear Information System (INIS)

Renato T, Banaga

1998-01-01

The Philippine Research Reactor (PRR-1) is the only research reactor in the Philippines. It was acquired through the Bilateral Agreement with the United States of America. The General Electric (G.E.) supplied PRR-1 first become operational in 1963 and used MTR plate type fuel. The original one-megawatt G.E. reactor was shutdown and converted into a 3 MW TRIGA PULSING REACTOR in 1984. The conversion includes the upgrading of the cooling system, replacement of new reactor coolant pumps, heat exchanger, cooling tower, replacement of new nuclear instrumentation and standard TRIGA console, TRIGA fuel supplied by General Atomic (G.A.). Philippine Nuclear Research Institute (PNRI) provided the old reactor, did the detailed design of the new cooling system, provided the new non-nuclear instrumentation and electrical power supply system and performed all construction, installation and modification work on site. The TRIGA conversion fuel is contained in a shrouded 4-rod cluster which fit into the original grid plate. The new fuel is a E 1 -U-Z 1 -H 1.6 TRIGA fuel, has a 20% wt Uranium loading with 19.7% U-235 enrichment and about 0.5 wt % Erbium. The Start-up, calibration and Demonstration of Pulsing and Full Power Operation were completed during a three week start-up phase which were performed last March 1968. A few days after, a leak in the pool liner was discovered. The reactor was shutdown again for repair and up to present the reactor is still in the process of rehabilitation. This paper will describe the rehabilitation/upgrading done on the PRR-1 since 1988 up to present. (author)

HTR-2002: Proceedings of the conference on high temperature reactors

International Nuclear Information System (INIS)

2002-01-01

High temperature reactors are considered as future inherently safe and efficient energy sources. The presentations covered all the relevant aspects of the existing HTGRs and/or helium cooled pebble bed reactors. They were sorted into 7 sessions: HTR Projects and Programmes; Fuel and Fuel Cycle; Physics and Neutronics; Thermohydraulic Calculation; Engineering, Design and Applications; Materials and Components; Safety and Licensing

Nuclear reactors. Introduction

International Nuclear Information System (INIS)

Boiron, P.

1997-01-01

This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)

Burnout experiments with 6 x 6, 8 x 8 and 7 x 7 rod bundle test sections using freon as model fluid

International Nuclear Information System (INIS)

Fulfs, H.; Katsaounis, A.; Minden, C.v.

1976-01-01

This paper reports on burnout experiments at staedy state condition using Freon12 as model fluid. The experiments were carried out with three test sections with 6 x 6, 8 x 8 and 7 x 7 rod bundles. The axial flux distribution of the rods is either constant or reactor like. The transformed measured points using STEVENS and BOURE scaling factors to equivalent water conditions respectively, were compared to the burnout correlation W3 using the reactor layout program DYNAMIT. The DYNAMIT code is a thermohydraulic lay-out reactor program without consideration of mixing flow between the subchannels. (orig.) [de

Nuclear reactor, reactor core thereof, and device for constituting the reactor

International Nuclear Information System (INIS)

Takiyama, Masashi.

1994-01-01

A reactor core is constituted by charging coolants (light water) in a reactor pressure vessel and distributing fuel assemblies, reflecting material sealing pipes, moderator (heavy water and helium gas) sealing pipes, and gas sealing pipes therein. A fuel guide tube is surrounded by a cap and the gap therebetween is made hollow and filled with coolant steams. The cap is supported by a baffle plate. The moderator sealing pipe is disposed in a flow channel of coolants in adjacent with the cap. The position of the moderator sealing tube in the reactor core is controlled by water stream from a hydraulic pump with a guide tube extending below the baffle plate being as a guide. Then, the position of the moderator sealing tube is varied to conduct power control, burnup degree compensation, and reactor shut down. With such procedures, moderator cooling facility is no more necessary to simplify the structure. Further, heat generated from the moderator is transferred to the coolants thereby improving heat efficiency of the reactor. (I.N.)