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.)

Steam generator tube performance: experience with water-cooled nuclear power reactors during 1978

International Nuclear Information System (INIS)

Tatone, O.S.; Pathania, R.S.

1980-02-01

The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1978. Tube failures occurred at 31 of the 86 reactors surveyed. Causes of these failures and procedures designed to deal with them are described. A dramatic decrease in the number of tubes plugged was evident in 1978 compared to the previous year. This is attributed to diligent application of techniques developed from in-plant experience and research and development programs over the past several years. (auth)

Steam generator tube performance: experience with water-cooled nuclear power reactors during 1977

International Nuclear Information System (INIS)

Pathania, R.S.; Tatone, O.S.

1979-02-01

The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1977. Failures were reported in 34 of the 79 reactors surveyed. Causes of these failures and inspection and repair procedures designed to deal with them are presented. Although corrosion remained the leading cause of tube failures, specific mechanisms have been identified and methods of dealing with them developed. These methods are being applied and should lead to a reduction of corrosion failures in future. (author)

Systems design of direct-cycle supercritical-water-cooled fast reactors

International Nuclear Information System (INIS)

Oka, Yoshiaki; Koshizuka, Seiichi; Jevremovic, Tatjana; Okano, Yashushi

1995-01-01

The system design of a direct-cycle supercritical-water-cooled fast reactor is presented. The supercritical water does not exhibit a change of phase. the recirculation system, steam separator, and dryer of a boiling water reactor (BWR) are unnecessary. Roughly speaking, the reactor pressure vessel and control rods are similar to those of a pressurized water reactor, the containment and emergency core cooling system are similar to a BWR, and the balance of plant is similar to a supercritical-pressure fossil-fired power plant (FPP). the electric power of the fast converter is 1,508 MW(electric). The number of coolant loops is only two because of the high coolant enthalpy. Containment volume is much reduced. The thermal efficiency is improved 24% over a BWR. The coolant void reactivity is negative by placing thin zirconium-hydride layers between seeds and blankets. The power costs would be much reduced compared with those of a light water reactor (LWR) and a liquid-metal fast breeder reactor. The concept is based on the huge amount of experience with the water coolant technology of LWRs and FPPs. The oxidation of stainless steel cladding is avoided by adopting a much lower coolant temperature than that of the FPP

Performance of water cooled nuclear power reactor fuels in India – Defects, failures and their mitigation

International Nuclear Information System (INIS)

Ganguly, Chaitanyamoy

2015-01-01

Water cooled and moderated nuclear power reactors account for more than 95% of the operating reactors in the world today. Light water reactors (LWRs) consisting of pressurized water reactor (PWR), their Russian counterpart namely VVER and boiling water reactor (BWR) will continue to dominate the nuclear power market. Pressurized heavy water reactor (PHWR), also known as CANDU, is the backbone of the nuclear power program in India. Updates on LWR and PHWR fuel performance are being periodically published by IAEA, OECD-NEA and the World Nuclear Association (WNA), highlighting fuel failure rate and the mitigation of fuel defects and failures. These reports clearly indicate that there has been significant improvement in in – pile fuel performance over the years and the present focus is to achieve zero fuel failure in high burn up and high performance fuels. The present paper summarizes the status of PHWR and LWR fuel performance in India, highlighting the manufacturing and the related quality control and inspection steps that are being followed at the PHWR fuel fabrication plant in order to achieve zero manufacturing defect which could contribute to achieving zero in – pile failure rate in operating and upcoming PHWR units in India. (author)

International conference on opportunities and challenges for water cooled reactors in the 21. century. PowerPoint presentations

International Nuclear Information System (INIS)

2009-01-01

Water Cooled Reactors have been the keystone of the nuclear industry in the 20th Century. As we move into the 21st Century and face new challenges such as the threat of climate change or the large growth in world energy demand, nuclear energy has been singled out as one of the sources that could substantially and sustainably contribute to power the world. As the nuclear community worldwide looks into the future with the development of advanced and innovative reactor designs and fuel cycles, it becomes important to explore the role Water Cooled Reactors (WCRs) will play in this future. To support the future role of WCRs, substantial design and development programmes are underway in a number of Member States to incorporate additional technology improvements into advanced nuclear power plants (NPPs) designs. One of the key features of advanced nuclear reactor designs is their improved safety due to a reduction in the probability and consequences of accidents and to an increase in the operator time allowed to better assess and properly react to abnormal events. A systematic approach and the experience of many years of successful operation have allowed designers to focus their design efforts and develop safer, more efficient and more reliable designs, and to optimize plant availability and cost through improved maintenance programs and simpler operation and inspection practices. Because many of these advanced WCR designs will be built in countries with no previous nuclear experience, it is also important to establish a forum to facilitate the exchange of information on the infrastructure and technical issues associated with the sustainable deployment of advanced nuclear reactors and its application for the optimization of maintenance of operating nuclear power plants. This international conference seeks to be all-inclusive, bringing together the policy, economic and technical decision-makers and the stakeholders in the nuclear industry such as operators, suppliers

Steam generator tube performance: experience with water-cooled nuclear power reactors during 1983 and 1984

International Nuclear Information System (INIS)

Tatone, O.S.; Meindl, P.; Taylor, G.F.

1986-06-01

A review of the performance of steam generator tubes in water-cooled nuclear power reactors showed that tubes were plugged at 47 (35.6%) of the reactors in 1983 and at 63 (42.6%) of the reactors during 1984. In 1983 and 1984 3291 and 3335 tubes, respectively, were removed from service, about the same as in 1982. The leading causes assigned to tube failure were stress corrosion cracking from the primary side and stress corrosion cracking or intergranular attack from the secondary side. In addition 5668 tubes were repaired for further service by installation of internal sleeves. Most of these were believed to have deteriorated by one of the above mechanisms or by pitting. There is a continuing trend towards high-integrity condenser tube materials at sites cooled by brackish or sea water. 31 refs

Steam generator tube failures: experience with water-cooled nuclear power reactors during 1976

International Nuclear Information System (INIS)

Tatone, O.S.; Pathania, R.S.

1978-02-01

A survey was conducted of experience with steam generator tubes at nuclear power stations during 1976. Failures were reported at 25 out of 68 water-cooled reactors. The causes of these failures and the repair and inspection procedures designed to cope with them are summarized. Examination of the data indicates that corrosion was the major cause of steam generator tube failures. Improvements are needed in steam generator design, condenser integrity and secondary water chemistry control. (author)

Natural Circulation Phenomena and Modelling for Advanced Water Cooled Reactors

International Nuclear Information System (INIS)

2012-03-01

The role of natural circulation in advanced water cooled reactor design has been extended with the adoption of passive safety systems. Some designs utilize natural circulation to remove core heat during normal operation. Most passive safety systems used in evolutionary and innovative water cooled reactor designs are driven by natural circulation. The use of passive systems based on natural circulation can eliminate the costs associated with the installation, maintenance and operation of active systems that require multiple pumps with independent and redundant electric power supplies. However, considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to ensure that the systems perform their intended functions. Several IAEA Member States with advanced reactor development programmes are actively conducting investigations of natural circulation to support the development of advanced water cooled reactor designs with passive safety systems. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, in 2004 the IAEA initiated a coordinated research project (CRP) on Natural Circulation Phenomena, Modelling and Reliability of Passive Systems that Utilize Natural Circulation. Three reports were published within the framework of this CRP. The first report (IAEA-TECDOC-1474) contains the material developed for the first IAEA training course on natural circulation in water cooled nuclear power plants. The second report (IAEA-TECDOC-1624) describes passive safety systems in a wide range of advanced water cooled nuclear power plant designs, with the goal of gaining insights into system design, operation and reliability. This third, and last, report summarizes the research studies completed by participating institutes during the CRP period.

Emergency cooling of presurized water reactor

International Nuclear Information System (INIS)

Sykora, D.

1981-01-01

The method described of emergency core cooling in the pressurized water reactor is characterized by the fact that water is transported to the disturbed primary circuit or direct to the reactor by the action of the energy and mass of the steam and/or liquid phase of the secondary circuit coolant, which during emergency core cooling becomes an emergency cooling medium. (B.S.)

Electrochemistry of Water-Cooled Nuclear Reactors

International Nuclear Information System (INIS)

Dgiby Macdonald; Mirna Urquidi-Macdonald; John Mahaffy; Amit Jain Han Sang Kim; Vishisht Gupta; Jonathan Pitt

2006-01-01

This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or ''radiation fields'' around the primary loop and the vessel, as a function of the operating parameters and the water chemistry

Occupational radiation exposure at light water cooled power reactors. Annual report, 1977

International Nuclear Information System (INIS)

Peck, L.J.

1979-04-01

This report presents an updated compilation of occupational radiation exposures at commercial light water cooled nuclear power reactors (LWRs) for the years 1969 through 1977. The information contained in this document was derived from reports submitted to the United States Nuclear Regulatory Commission in accordance with requirements of individual plant Technical Specifications, and in accordance with Part 20.407 of Title 10, Chapter 1, Code of Federal Regulations (10 CFR Part 20.407). An additional 4 LWRs completed a full calendar year of commercial operation for the first time in 1977. This report now encompasses data from 57 commercially operating U.S. nuclear power plants. The number of personnel monitored at LWRs increased approximately 10% in 1977, and the average collective dose to personnel (man-rems per reactor-year) increased 14% over the 1976 average. The average number of personnel receiving measurable exposure per reactor increased 11%, and the average exposure per individual in 1977 was 0.8 rem per person

Steam-generator tube performance: world experience with water-cooled nuclear power reactors during 1978

International Nuclear Information System (INIS)

Tatone, O.S.; Pathania, R.S.

1980-01-01

The performance of steam-generator tubes in water-cooled nuclear power reactors during 1978 is reviewed. Tube failures occurred at 31 of the 86 reactors surveyed. The causes of these failures and the procedures designed to deal with them are described. The number of tubes plugged has decreased dramatically in 1978 compared to the previous year. This is attributed to the diligent application of techniques developed through in-plant experience and research and development programs over the past several years

Improving activity transport models for water-cooled nuclear power reactors

Energy Technology Data Exchange (ETDEWEB)

Burrill, K.A

2001-08-01

Eight current models for describing radioactivity transport and radiation field growth around water-cooled nuclear power reactors have been reviewed and assessed. A frequent failing of the models is the arbitrary nature of the determination of the important processes. Nearly all modelers agree that the kinetics of deposition and release of both dissolved and particulate material must be described. Plant data must be used to guide the selection and development of suitable improved models, with a minimum of empirically-based rate constraints being used. Limiting case modelling based on experimental data is suggested as a way to simplify current models and remove their subjectivity. Improved models must consider the recent change to 'coordinated water chemistry' that appears to produce normal solubility behaviour for dissolved iron throughout the fuel cycle in PWRs, but retrograde solubility remains for dissolved nickel. Profiles are suggested for dissolved iron and nickel concentrations around the heat transport system in CANDU reactors, which operate nominally at constant chemistry, i.e., pH{sub T} constant with time, and which use carbon steel isothermal piping. These diagrams are modified for a CANDU reactor with stainless steel piping, in order to show the changes expected. The significance of these profiles for transport in PWRs is discussed for further model improvement. (author)

Guide to power reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1959-07-15

The IAEA's major first scientific publication is the Directory of Power Reactors now in operation or under construction in various parts of the world. The purpose of the directory is to present important details of various power projects in such a way as to provide a source of easy reference for anyone interested in the development of the peaceful uses of atomic energy, either at the technical or management level. Six pages have been devoted to each reactor the first of which contains general information, reactor physics data and information about the core. The second and third contain sketches of the fuel element or of the fuel element assembly, and of the horizontal and vertical sections of the reactor. On the fourth page information is grouped under the following heads: fuel element, core heat transfer, control, reactor vessel and over-all dimensions, and fluid flow. The fifth page shows a simplified flow diagram, while the sixth provides information on reflector and shielding, containment and turbo generator. Some information has also been given, when available, on cost estimates and operating staff requirements. Remarks and a bibliography constitute the last part of the description of each reactor. Reactor projects included in this directory are pressurized light water cooled power reactors. Boiling light water cooled power reactors, heavy water cooled power reactors, gas cooled power reactors, organic cooled power reactors liquid metal cooled power reactors and liquid metal cooled power reactors

Cooling of nuclear power stations with high temperature reactors and helium turbine cycles

International Nuclear Information System (INIS)

Foerster, S.; Hewing, G.

1977-01-01

On nuclear power stations with high temperature reactors and helium turbine cycles (HTR-single circuits) the residual heat from the energy conversion process in the primary and intermediate coolers is removed from cycled gas, helium. Water, which is circulated for safety reasons through a closed circuit, is used for cooling. The primary and intermediate coolers as well as other cooling equipment of the power plant are installed within the reactor building. The heat from the helium turbine cycle is removed to the environment most effectively by natural draught cooling towers. In this way a net plant efficiency of about 40% is attainable. The low quantities of residual heat thereby produced and the high (in comparison with power stations with steam turbine cycles) cooling agent pressure and cooling water reheat pressure in the circulating coolers enable an economically favourable design of the overall 'cold end' to be expected. In the so-called unit range it is possible to make do with one or two cooling towers. Known techniques and existing operating experience can be used for these dry cooling towers. After-heat removal reactor shutdown is effected by a separate, redundant cooling system with forced air dry coolers. The heat from the cooling process at such locations in the power station is removed to the environment either by a forced air dry cooling installation or by a wet cooling system. (orig.) [de

Advanced water-cooled reactor technologies. Rationale, state of progress and outlook

International Nuclear Information System (INIS)

Anon.

1989-01-01

Eighty per cent of the world's power reactors are water cooled and moderated. Many improvements in their design and operation have been implemented since the first such reactor started commercial operation in 1957. This report addresses the safety, environmental and economic rationales for further improvements, as well as their relevance to currently operating water reactors

Thermal-hydraulic R and D infrastructure for water cooled reactors of the Indian nuclear power program

International Nuclear Information System (INIS)

Vijayan, P.K.; Jain, V.; Saha, D.; Sinha, R.K.

2009-01-01

R and D has been the critical ingredient of Indian Nuclear Power Program from the very inception. Approach to R and D infrastructure has been closely associated with the three-stage nuclear power program that was crafted on the basis of available resources and technology in the short-term and energy security in the long-term. Early R and D efforts were directed at technologies relevant to Pressurized Heavy Water Reactors (PHWRs) which are currently the mainstay of Indian nuclear power program. Lately, the R and D program has been steered towards the design and development of advanced and innovative reactors with the twin objective of utilization of abundant thorium and to meet the future challenges to nuclear power such as enhanced safety and reliability, better economy, proliferation resistance etc. Advanced Heavy Water Reactor (AHWR) is an Indian innovative reactor currently being developed to realize the above objectives. Extensive R and D infrastructure has been created to validate the system design and various passive concepts being incorporated in the AHWR. This paper provides a brief review of R and D infrastructure that has been developed at Bhabha Atomic Research Centre for thermal-hydraulic investigations for water-cooled reactors of Indian nuclear power program. (author)

Development Project of Supercritical-water Cooled Power Reactor

International Nuclear Information System (INIS)

Kataoka, K.; Shiga, S.; Moriya, K.; Oka, Y.; Yoshida, S.; Takahashi, H.

2002-01-01

A Supercritical-water Cooled Power Reactor (SCPR) development project (Feb. 2001- Mar. 2005) is being performed by a joint team consisting of Japanese universities and nuclear venders with a national fund. The main objective of this project is to provide technical information essential to demonstration of SCPR technologies through concentrating three sub-themes: 'plant conceptual design', 'thermohydraulics', and 'material and water chemistry'. The target of the 'plant conceptual design sub-theme' is simplify the whole plant systems compared with the conventional LWRs while achieving high thermal efficiency of more than 40 % without sacrificing the level of safety. Under the 'thermohydraulics sub-theme', heat transfer characteristics of supercritical-water as a coolant of the SCPR are examined experimentally and analytically focusing on 'heat transfer deterioration'. The experiments are being performed using fron-22 for water at a fossil boiler test facility. The experimental results are being incorporated in LWR analytical tools together with an extended steam/R22 table. Under the 'material and water chemistry sub-theme', material candidates for fuel claddings and internals of the SCPR are being screened mainly through mechanical tests, corrosion tests, and simulated irradiation tests under the SCPR condition considering water chemistry. In particular, stress corrosion cracking sensitivity is being investigated as well as uniform corrosion and swelling characteristics. Influences of water chemistry on the corrosion product characteristics are also being examined to find preferable water condition as well as to develop rational water chemistry controlling methods. (authors)

Thermal and stability considerations for a supercritical water-cooled fast reactor during power-raising phase of plant startup

International Nuclear Information System (INIS)

Cai, Jiejin; Ishiwatari, Yuki; Oka, Yoshiaki; Ikejiri, Satoshi

2009-01-01

This paper describes thermal analyses and linear stability analyses of the Supercritical Water-cooled Fast Reactor with 'two-path' flow scheme during the power-raising phase of plant startup. For thermal consideration, the same criterion of the maximum cladding surface temperature (MCST) as applied to the normal operating condition is used. For thermal-hydraulic stability consideration, the decay ratio of 0.5 is applied, which is taken from BWRs. Firstly, we calculated the flow rate distribution among the parallel flow paths from the reactor vessel inlet nozzles to the mixing plenum below the core using a system analysis code. The parallel flow paths consist of the seed fuel assemblies cooled by downward flow, the blanket fuel assemblies cooled by downward flow and the downcomer. Then, the MCSTs are estimated for various reactor powers and feedwater flow rates with system analyses. The decay ratios are estimated with linear stability analyses. The available range of the reactor power and feedwater flow rate to satisfy the thermal and stability criteria is obtained. (author)

Cooling device for reactor container

International Nuclear Information System (INIS)

Arai, Kenji.

1996-01-01

Upon assembling a static container cooling system to an emergency reactor core cooling system using dynamic pumps in a power plant, the present invention provides a cooling device of lowered center of gravity and having a good cooling effect by lowering the position of a cooling water pool of the static container cooling system. Namely, the emergency reactor core cooling system injects water to the inside of a pressure vessel using emergency cooling water stored in a suppression pool as at least one water source upon loss of reactor coolant accident. In addition, a cooling water pool incorporating a heat exchanger is disposed at the circumference of the suppression pool at the outside of the container. A dry well and the heat exchanger are connected by way of steam supply pipes, and the heat exchanger is connected with the suppression pool by way of a gas exhaustion pipe and a condensate returning pipeline. With such a constitution, the position of the heat exchanger is made higher than an ordinary water level of the suppression pool. As a result, the emergency cooling water of the suppression pool water is injected to the pressure vessel by the operation of the reactor cooling pumps upon loss of coolant accident to cool the reactor core. (I.S.)

Condensation nuclear power plants with water-cooled graphite-moderated channel type reactors and advances in their development

International Nuclear Information System (INIS)

Boldyrev, V.M.; Mikhaj, V.I.

1985-01-01

Consideration is being given to results of technical and economical investigations of advisability of increasing unit power by elevating steam generating capacity as a result of inserting numerous of stereotype sectional structural elements of the reactor with similar thermodynamic parameters. It is concluded that construction of power units of condensation nuclear power plants with water-cooled graphite-moderated channel type reactors of 2400-3200 MWe and higher unit power capacity represents the real method for sharp growth of efficiency and labour productivity in power industry. It can also provide the required increase of the rate of putting electrogenerating powers into operation

Thermohydraulic relationships for advanced water cooled reactors

International Nuclear Information System (INIS)

2001-04-01

This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

Thermohydraulic relationships for advanced water cooled reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-04-01

This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

Gas cooled reactors

International Nuclear Information System (INIS)

Kojima, Masayuki.

1985-01-01

Purpose: To enable direct cooling of reactor cores thereby improving the cooling efficiency upon accidents. Constitution: A plurality sets of heat exchange pipe groups are disposed around the reactor core, which are connected by way of communication pipes with a feedwater recycling device comprising gas/liquid separation device, recycling pump, feedwater pump and emergency water tank. Upon occurrence of loss of primary coolants accidents, the heat exchange pipe groups directly absorb the heat from the reactor core through radiation and convection. Although the water in the heat exchange pipe groups are boiled to evaporate if the forcive circulation is interrupted by the loss of electric power source, water in the emergency tank is supplied due to the head to the heat exchange pipe groups to continue the cooling. Furthermore, since the heat exchange pipe groups surround the entire circumference of the reactor core, cooling is carried out uniformly without resulting deformation or stresses due to the thermal imbalance. (Sekiya, K.)

Development of an automated system of nuclear materials accounting for nuclear power stations with water-cooled, water-moderated reactors

International Nuclear Information System (INIS)

Babaev, N.S.

1981-06-01

The results of work carried out under IAEA Contract No. 2336/RB are described (subject: an automated system of nuclear materials accounting for nuclear power stations with water-cooled, water-moderated (VVER) reactors). The basic principles of an accounting system for this type of nuclear power plant are outlined. The general structure and individual units of the information computer program used to achieve automated accounting are described and instructions are given on the use of the program. A detailed example of its application (on a simulated nuclear power plant) is examined

Material Issues of Blanket Systems for Fusion Reactors - Compatibility with Cooling Water -

Science.gov (United States)

Miwa, Yukio; Tsukada, Takashi; Jitsukawa, Shiro

Environmental assisted cracking (EAC) is one of the material issues for the reactor core components of light water power reactors(LWRs). Much experience and knowledge have been obtained about the EAC in the LWR field. They will be useful to prevent the EAC of water-cooled blanket systems of fusion reactors. For the austenitic stainless steels and the reduced-activation ferritic/martensitic steels, they clarifies that the EAC in a water-cooled blanket does not seem to be acritical issue. However, some uncertainties about influences on water temperatures, water chemistries and stress conditions may affect on the EAC. Considerations and further investigations elucidating the uncertainties are discussed.

Feasibility study of self sustaining capability on water cooled thorium reactors for different power reactors

International Nuclear Information System (INIS)

Permana, S.; Takaki, N.; Sekimoto, H.

2007-01-01

Thorium fuel cycle can maintain the sustainable system of the reactor for self sustaining system for future sustainable development in the world. Some characteristics of thorium cycle show some advantages in relation to higher breeding capability, higher performance of burn-up and more proliferation resistant. Several investigations was performed to improve the breeding capability which is essential for maintaining the fissile sustainability during reactor operation in thermal reactor such as Shippingport reactor and molten salt breeder reactor (MSBR) project. The preliminary study of breeding capability on water cooled thorium reactor has been investigated for various power output. The iterative calculation system is employed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of PIJ module of SRAC2000. In this calculation, 1238 fission products and 129 heavy nuclides are employed. In the cell calculation, 26 heavy metals and 66 fission products and 1 pseudo FP are employed. The employed nuclear data library was JENDL 3.2. The reactor is fueled by 2 33U-Th Oxide and it has used the light water coolant as moderator. Some characteristics such as conversion ratio and void reactivity coefficient performances are evaluated for the systems. The moderator to fuel ratio (MFR) values and average burnups are studied for survey parameter. The parametric survey for different power outputs are employed from 10 MWt to 3000 MWt for evaluating the some characteristics of core size and leakage effects to the spectra profile, required enrichment, breeding capability, fissile inventory condition, and void reactivity coefficient. Different power outputs are employed in order to evaluate its effect to the required enrichment for criticality, breeding capability, void reactivity and fissile inventory accumulation. The obtained value of the conversion ratios is evaluated by using the equilibrium atom composition. The conversion ratio is employed based on the

Power flattening and reactivity suppression strategies for the Canadian supercritical water reactor concept

International Nuclear Information System (INIS)

McDonald, M.; Colton, A.; Pencer, J.

2015-01-01

The Canadian supercritical water-cooled reactor (SCWR) is a conceptual heavy water moderated, supercritical light water cooled pressure tube reactor. In contrast to current heavy water power reactors, the Canadian SCWR will be a batch fuelled reactor. Associated with batch fuelling is a large beginning-of-cycle excess reactivity. Furthermore, radial power peaking arising as a consequence of batch refuelling must be mitigated in some way. In this paper, burnable neutron absorber (BNA) added to fuel and absorbing rods inserted into the core are considered for reactivity management and power flattening. A combination of approaches appears adequate to reduce the core radial power peaking, while also providing reactivity suppression. (author)

Improving economics and safety of water cooled reactors. Proven means and new approaches

International Nuclear Information System (INIS)

2002-05-01

Nuclear power plants (NPPs) with water cooled reactors [either light water reactors (LWRs) or heavy water reactors (HWRs)] constitute the large majority of the currently operating plants. Water cooled reactors can make a significant contribution to meeting future energy needs, to reducing greenhouse gas emissions, and to energy security if they can compete economically with fossil alternatives, while continuing to achieve a very high level of safety. It is generally agreed that the largest commercial barrier to the addition of new nuclear power capacity is the high capital cost of nuclear plants relative to other electricity generating alternatives. If nuclear plants are to form part of the future generating mix in competitive electricity markets, capital cost reduction through simplified designs must be an important focus. Reductions in operating, maintenance and fuel costs should also be pursued. The Department of Nuclear Energy of the IAEA is examining the competitiveness of nuclear power and the means for improving its economics. The objective of this TECDOC is to emphasize the need, and to identify approaches, for new nuclear plants with water cooled reactors to achieve competitiveness while maintaining high levels of safety. The cost reduction methods discussed herein can be implemented into plant designs that are currently under development as well as into designs that may be developed in the longer term. Many of the approaches discussed also generally apply to other reactor types (e.g. gas cooled and liquid metal cooled reactors). To achieve the largest possible cost reductions, proven means for reducing costs must be fully implemented, and new approaches described in this document should be developed and implemented. These new approaches include development of advanced technologies, increased use of risk-informed methods for evaluating the safety benefit of design features, and international consensus regarding commonly acceptable safety requirements that

Improving economics and safety of water cooled reactors. Proven means and new approaches

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-05-01

Nuclear power plants (NPPs) with water cooled reactors [either light water reactors (LWRs) or heavy water reactors (HWRs)] constitute the large majority of the currently operating plants. Water cooled reactors can make a significant contribution to meeting future energy needs, to reducing greenhouse gas emissions, and to energy security if they can compete economically with fossil alternatives, while continuing to achieve a very high level of safety. It is generally agreed that the largest commercial barrier to the addition of new nuclear power capacity is the high capital cost of nuclear plants relative to other electricity generating alternatives. If nuclear plants are to form part of the future generating mix in competitive electricity markets, capital cost reduction through simplified designs must be an important focus. Reductions in operating, maintenance and fuel costs should also be pursued. The Department of Nuclear Energy of the IAEA is examining the competitiveness of nuclear power and the means for improving its economics. The objective of this TECDOC is to emphasize the need, and to identify approaches, for new nuclear plants with water cooled reactors to achieve competitiveness while maintaining high levels of safety. The cost reduction methods discussed herein can be implemented into plant designs that are currently under development as well as into designs that may be developed in the longer term. Many of the approaches discussed also generally apply to other reactor types (e.g. gas cooled and liquid metal cooled reactors). To achieve the largest possible cost reductions, proven means for reducing costs must be fully implemented, and new approaches described in this document should be developed and implemented. These new approaches include development of advanced technologies, increased use of risk-informed methods for evaluating the safety benefit of design features, and international consensus regarding commonly acceptable safety requirements that

Heavy water moderated gas-cooled reactors

International Nuclear Information System (INIS)

Bailly du Bois, B.; Bernard, J.L.; Naudet, R.; Roche, R.

1964-01-01

France has based its main effort for the production of nuclear energy on natural Uranium Graphite-moderated gas-cooled reactors, and has a long term programme for fast reactors, but this country is also engaged in the development of heavy water moderated gas-cooled reactors which appear to present the best middle term prospects. The economy of these reactors, as in the case of Graphite, arises from the use of natural or very slightly enriched Uranium; heavy water can take the best advantages of this fuel cycle and moreover offers considerable development potential because of better reactor performances. A prototype plant EL 4 (70 MW) is under construction and is described in detail in another paper. The present one deals with the programme devoted to the development of this reactor type in France. Reasons for selecting this reactor type are given in the first part: advantages and difficulties are underlined. After reviewing the main technological problems and the Research and Development carried out, results already obtained and points still to be confirmed are reported. The construction of EL 4 is an important step of this programme: it will be a significant demonstration of reactor performances and will afford many experimentation opportunities. Now the design of large power reactors is to be considered. Extension and improvements of the mechanical structures used for EL 4 are under study, as well as alternative concepts. The paper gives some data for a large reactor in the present state of technology, as a result from optimization studies. Technical improvements, especially in the field of materials could lead to even more interesting performances. Some prospects are mentioned for the long run. Investment costs and fuel cycles are discussed in the last part. (authors) [fr

Status of advanced technology and design for water cooled reactors: Heavy water reactors

International Nuclear Information System (INIS)

1989-07-01

In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of the IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors, has been undertaken to document the major current activities and trends of technological improvement and development for future water reactors. Part I of the report dealing with Light Water Reactors (LWRs) was published in 1988 (IAEA-TECDOC-479). Part II of the report covers Heavy Water Reactors (HWRs) and has now been prepared. This report is based largely upon submissions from Member States. It has been supplemented by material from the presentations at the IAEA Technical Committee and Workshop on Progress in Heavy Water Reactor Design and Technology held in Montreal, Canada, December 6-9, 1988. It is hoped that this part of the report, containing the status of advanced heavy water reactor technology up to 1988 and ongoing development programmes will aid in disseminating information to Member States and in stimulating international cooperation. Refs, figs and tabs

Supercritical-pressure light water cooled reactors

CERN Document Server

Oka, Yoshiaki

2014-01-01

This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

Stability analysis of supercritical-pressure light water-cooled reactor in constant pressure operation

International Nuclear Information System (INIS)

Suhwan, JI; Shirahama, H.; Koshizuka, S.; Oka, Y.

2001-01-01

The purpose of this study is to evaluate the thermal-hydraulic and the thermal-nuclear coupled stabilities of a supercritical pressure light water-cooled reactor. A stability analysis code at supercritical pressure is developed. Using this code, stabilities of full and partial-power reactor operating at supercritical pressure are investigated by the frequency-domain analysis. Two types of SCRs are analyzed; a supercritical light water reactor (SCLWR) and a supercritical water-cooled fast reactor (SCFR). The same stability criteria as Boiling Water Reactor are applied. The thermal-hydraulic stability of SCLWR and SCFR satisfies the criteria with a reasonable orifice loss coefficient. The decay ratio of the thermal-nuclear coupled stability in SCFR is almost zero because of a small coolant density coefficient of the fast reactor. The evaluated decay ratio of the thermal-nuclear coupled stability is 3,41 ∼ 10 -V at 100% power in SCFR and 0,028 at 100% power in SCLWR. The sensitivity is investigated. It is found that the thermal-hydraulic stability is sensitive to the mass flow rate strongly and the thermal-nuclear coupled stability to the coolant density coefficient. The bottom power peak distribution makes the thermal-nuclear stability worse and the thermal-nuclear stability better. (author)

Supercritical-pressure, once-through cycle light water cooled reactor concept

International Nuclear Information System (INIS)

Oka, Yoshiaki; Koshizuka, Seiichi

2001-01-01

The purpose of the study is to develop new reactor concepts for the innovation of light water reactors (LWR) and fast reactors. Concept of the once-through coolant cycle, supercritical-pressure light water cooled reactor was developed. Major aspects of reactor design and safety were analysed by the computer codes which were developed by ourselves. It includes core design of thermal and fast reactors, plant system, safety criteria, accident and transient analysis, LOCA, PSA, plant control, start up and stability. High enthalpy rise as supercritical boiler was achieved by evaluating the cladding temperature directly during transients. Fundamental safety principle of the reactor is monitoring coolant flow rate instead of water level of LWR. The reactor system is compact and simple because of high specific enthalpy of supercritical water and the once-through cycle. The major components are similar to those of LWR and supercritical thermal plant. Their temperature are within the experiences in spite of the high outlet coolant temperature. The reactor is compatible with tight fuel lattice fast reactor because of the high head pumps and low coolant flow rate. The power rating of the fast reactor is higher than the that of thermal reactor because of the high power density. (author)

Steam generator tube performance: world experience with water-cooled nuclear power reactors during 1979

International Nuclear Information System (INIS)

Tatone, O.S.; Pathania, R.S.

1981-01-01

The performance of steam generator tubes in water-cooled nuclear power reactors is reviewed for 1979. Tube failures occurred at 38 of the 93 reactors surveyed. The causes of these failures and the procedures designed to deal with them are described. The defect rate, although higher than that in 1978, was still lower than the rates of the two previous years. Methods being employed to detect defects include the increased use of multifrequency eddy-current testing and a trend to full-length inspection of all tubes. To reduce the incidence of tube failure by corrosion, plant operators are turning to full-flow condensate demineralization and more leak-resistant condenser tubes. 10 tables

Effect of dc-power-system reliability on reactor-shutdown cooling

International Nuclear Information System (INIS)

Kolaczkowski, A.M.; Baranowsky, P.W.; Hickman, J.W.

1981-01-01

The DC power systems in a nuclear power plant provide control and motive power to valves, instrumentation, emergency diesel generators, and many other components and systems during all phases of plant operation including abnormal shutdowns and accident situations. A specific area of concern is the adequacy of the minimum design requirements for DC power systems, particularly with regard to multiple and common cause failures. This concern relates to the application of the single failure criterion for assuring a reliable DC power supply which may be required for the functionability of shutdown cooling systems. The results are presented of a reliability based study performed to assess the adequacy of DC power supply design requirements for currently operating light water reactors with particular attention to shutdown cooling requirements

Hydrogen in water-cooled nuclear power reactors

International Nuclear Information System (INIS)

1992-01-01

The Commission of the European Community (CEC) and the International Atomic Energy Agency (IAEA) decided in 1989 to update the state of the art concerning hydrogen in water cooled nuclear power reactors by commissioning a report which would review, all the available information to-date and make recommendations for the future. This joint report was prepared by committees formed by the IAEA and by the CEC. The aim of this report is to review the current understanding on the areas in which the research on hydrogen in LWR is conventionally presented, taking into account the results of the latest reported research developments. The main reactions through which hydrogen is produced are assessed together with their timings. An estimation of the amount of hydrogen produced by each reaction is given, in order to reckon their relative contribution to the hazard. An overview is then given of the state of knowledge of the most important phenomena taking place during its transport from the place of production and the phenomena which control the hydrogen combustion and the consequences of combustion under various conditions. Specific research work is recommended in each sector of the presented phenomena. The last topics reviewed in this report are the hydrogen detection and the prevent/mitigation of pressure and temperature loads on containment structures and structures and safety related equipment caused by hydrogen combustion

Delayed gamma power measurement for sodium-cooled fast reactors

Energy Technology Data Exchange (ETDEWEB)

Coulon, R., E-mail: romain.coulon@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Normand, S., E-mail: stephane.normand@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Ban, G., E-mail: ban@lpccaen.in2p3.f [ENSICAEN, 6 Boulevard Marechal Juin, F-14050 Caen Cedex 4 (France); Barat, E.; Montagu, T.; Dautremer, T. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Brau, H.-P. [ICSM, Centre de Marcoule, BP 17171 F-30207 Bagnols sur Ceze (France); Dumarcher, V. [AREVA NP, SET, F-84500 Bollene (France); Michel, M.; Barbot, L.; Domenech, T.; Boudergui, K.; Bourbotte, J.-M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Jousset, P. [CEA, LIST, Departement des Capteurs, du Signal et de l' Information, F-91191 Gif-sur-Yvette (France); Barouch, G.; Ravaux, S.; Carrel, F. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Saurel, N. [CEA, DAM, Laboratoire Mesure de Dechets et Expertise, F-21120 Is-sur-Tille (France); Frelin-Labalme, A.-M.; Hamrita, H. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France)

2011-01-15

Graphical abstract: Display Omitted Research highlights: {sup 20}F and {sup 23}Ne tagging agents are produced by fast neutron flux. {sup 20}F signal has been measured at the SFR Phenix prototype. A random error of only 3% for an integration time of 2 s could be achieved. {sup 20}F and {sup 23}Ne power measurement has a reduced temperature influence. Burn-up impact could be limited by simultaneous {sup 20}F and {sup 23}Ne measurement. - Abstract: Previous works on pressurized water reactors show that the nitrogen 16 activation product can be used to measure thermal power. Power monitoring using a more stable indicator than ex-core neutron measurements is required for operational sodium-cooled fast reactors, in order to improve their economic efficiency at the nominal operating point. The fluorine 20 and neon 23 produced by (n,{alpha}) and (n,p) capture in the sodium coolant have this type of convenient characteristic, suitable for power measurements with low build-up effects and a potentially limited temperature, flow rate, burn-up and breeding dependence. This method was tested for the first time during the final tests program of the French Phenix sodium-cooled fast reactor at CEA Marcoule, using the ADONIS gamma pulse analyzer. Despite a non-optimal experimental configuration for this application, the delayed gamma power measurement was pre-validated, and found to provide promising results.

A gas-cooled reactor surface power system

International Nuclear Information System (INIS)

Lipinski, R.J.; Wright, S.A.; Lenard, R.X.; Harms, G.A.

1999-01-01

A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars. copyright 1999 American Institute of Physics

A gas-cooled reactor surface power system

International Nuclear Information System (INIS)

Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

1999-01-01

A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars

A Gas-Cooled Reactor Surface Power System

Energy Technology Data Exchange (ETDEWEB)

Harms, G.A.; Lenard, R.X.; Lipinski, R.J.; Wright, S.A.

1998-11-09

A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life- cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitide clad in Nb 1 %Zr, which has been extensively tested under the SP-I 00 program The fiel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fbel and stabilizing the geometty against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality cannot occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

Concept of safe tank-type water cooled and moderated reactor with HTGR microparticle fuel compacts

International Nuclear Information System (INIS)

Gol'tsev, A.O.; Kukharkin, N.E.; Mosevitskij, I.S.; Ponomarev-Stepnoj, N.N.; Popov, S.V.; Udyanskij, Yu.N.; Tsibul'skij, V.F.

1993-01-01

Concept of safe tank-type water-cooled and moderated reactor on the basis of HTGR fuel microparticles which enable to avoid environment contamination with radioactive products under severe accidents, is proposed. Results of neutron-physical and thermal-physical studies of water cooled and moderated reactor with HTGR microparticle compacts are presented. Characteristics of two reactors with thermal power of 500 and 1500 MW are indicated within the concept frames. The reactor behaviour under severe accident connected with complete loss of water coolant is considered. It is shown that under such an accident the fission products release from fuel microparticles does not occur

Indian experience with radionuclide transport, deposition and decontamination in water-cooled nuclear power reactors

International Nuclear Information System (INIS)

Narasimhan, S.V.; Das, P.C.; Lawrence, D.A.; Mathur, P.K.; Venkateswarlu, K.S.

1983-01-01

The present generation of water-cooled nuclear reactors uses construction materials chosen with utmost care so that minimum corrosion occurs during the life of the reactor. As interaction between the primary coolant and the construction materials is unavoidable, the coolant is chemically treated to achieve maximum compatibility. First measurements of the chemical and radiochemical composition of the crud present on the in-core and out-of-core primary heat transport system surfaces of a pressurized heavy-water-moderated and cooled reactor (PHWR) are given; then experience in India in the development of a low temperature, one-stage decontaminating formulation for chemical decontamination of the radioactive deposits formed on stainless steel surfaces under BWR conditions is discussed. The effect of the magnitude of the transients in parameters such as reactor power, system temperature, dissolved oxygen content in the coolant, etc. on the nature and migration behaviour of primary heat transport system crud in a PHWR is described. Contributions to radioactive sources and insoluble crud from different primary heat transport system materials are identified and correlated with reactor operations in a PHWR. Man-rem problems faced by nuclear reactors, especially during off-line maintenance, stress the need for reducing the deposited radioactive sources from system surfaces which would otherwise be accessible. Laboratory and on-site experimentation was carried out to effect chemical decontamination on the radioactive deposits formed on the stainless steel surfaces under BWR conditions. Both the reducing and oxidizing formulations were subsequently used in a small-scale, in-plant trial in the clean-up system of a BWR. More than 85% of the deposited 60 Co activity was found to have been removed by the oxidizing formulation. Efforts to develop a decontaminating mixture containing a reducing agent with the help of a circulating loop are in progress in the laboratory. (author)

Method of injecting cooling water in emergency core cooling system (ECCS) of PWR type reactor

International Nuclear Information System (INIS)

Sobajima, Makoto; Adachi, Michihiro; Tasaka, Kanji; Suzuki, Mitsuhiro.

1979-01-01

Purpose: To provide a cooling water injection method in an ECCS, which can perform effective cooling of the reactor core. Method: In a method of injecting cooling water in an ECCS as a countermeasure against a rupture accident of a pwr type reactor, cooling water in the first pressure storage injection system is injected into the upper plenum of the reactor pressure vessel at a set pressure of from 50 to 90 atg. and a set temperature of from 80 to 200 0 C, cooling water in the second pressure storage injection system is injected into the lower plenum of the reactor pressure vessel at a pressure of from 25 to 60 atg. which is lower than the set pressure and a temperature less than 60 0 C, and further in combination with these procedures, cooling water of less than 60 0 C is injected into a high-temperature side piping, in the high-pressure injection system of upstroke of 100 atg. by means of a pump and the low-pressure injection system of upstroke of 20 atg. also by means of a pump, thereby cooling the reactor core. (Aizawa, K.)

Method of operating water cooled reactor with blanket

International Nuclear Information System (INIS)

Suzuki, Katsuo.

1988-01-01

Purpose: To increase the production amount of fissionable plutonium by increasing the burnup degree of blanket fuels in a water cooled reactor with blanket. Method: Incore insertion assemblies comprising water elimination rods, fertile material rods or burnable poison rods are inserted to those fuel assemblies at the central portion of the reactor core that are situated at the positions not inserted with control rods in the earlier half of the operation cycle, while the incore reactor insertion assemblies are withdrawn at the latter half of the operation cycle of a nuclear reactor. As a result, it is possible to increase the power share of the blanket fuels and increase the fuel burnup degree to thereby increase the production amount of fissionable plutonium. Furthermore, at the initial stage of the cycle, the excess reactivity of the reactor can be suppressed to decrease the reactivity control share on the control rod. At the final stage of the cycle, the excess reactivity of the reactor core can be increased to improve the cycle life. (Kamimura, M.)

Changes in water chemistry and primary productivity of a reactor cooling reservoir (Par Pond)

International Nuclear Information System (INIS)

Tilly, L.J.

1975-01-01

Water chemistry and primary productivity of a reactor cooling reservoir have been studied for 8 years. Initially the primary productivity increased sixfold, and the dissolved solids doubled. The dissolved-solids increase appears to have been caused by additions of makeup water from the Savannah River and by evaporative concentration during the cooling process. As the dissolved-solids concentrations and the conductivity of makeup water leveled off, the primary productivity stabilized. Major cation and anion concentrations generally followed total dissolved solids through the increase and plateau; however, silica concentrations declined steadily during the initial period of increased plankton productivity. Standing crops of net seston and centrifuge seston did not increase during this initial period. The collective data show the effects of thermal input to a cooling reservoir, illustrate the need for limnological studies before reactor siting, and suggest the possibility of using makeup-water additions to power reactor cooling basins as a reservoir management tool

Meltdown reactor core cooling facility

International Nuclear Information System (INIS)

Matsuoka, Tsuyoshi.

1992-01-01

The meltdown reactor core cooling facility comprises a meltdown reactor core cooling tank, a cooling water storage tank situates at a position higher than the meltdown reactor core cooling tank, an upper pipeline connecting the upper portions of the both of the tanks and a lower pipeline connecting the lower portions of them. Upon occurrence of reactor core meltdown, a high temperature meltdown reactor core is dropped on the cooling tank to partially melt the tank and form a hole, from which cooling water is flown out. Since the water source of the cooling water is the cooling water storage tank, a great amount of cooling water is further dropped and supplied and the reactor core is submerged and cooled by natural convection for a long period of time. Further, when the lump of the meltdown reactor core is small and the perforated hole of the meltdown reactor cooling tank is small, cooling water is boiled by the high temperature lump intruding into the meltdown reactor core cooling tank and blown out from the upper pipeline to the cooling water storage tank to supply cooling water from the lower pipeline to the meltdown reactor core cooling tank. Since it is constituted only with simple static facilities, the facility can be simplified to attain improvement of reliability. (N.H.)

European supercritical water cooled reactor

International Nuclear Information System (INIS)

Schulenberg, T.; Starflinger, J.; Marsault, P.; Bittermann, D.; Maraczy, C.; Laurien, E.; Lycklama a Nijeholt, J.A.; Anglart, H.; Andreani, M.; Ruzickova, M.; Toivonen, A.

2011-01-01

Highlights: → The HPLWR reactor design is an example of a supercritical water cooled reactor. → Cladding material tests have started but materials are not yet satisfactory. → Numerical heat transfer predictions are promising but need further validation. → The research project is most suited for nuclear education and training. - Abstract: The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 o C average core outlet temperature. It is designed and analyzed by a European consortium of 10 partners and 3 active supporters from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small fuel assemblies in boxes with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. Besides the scientific and technical progress, the HPLWR project turned out to be most successful in training the young generation of nuclear engineers

Water cooled nuclear reactor

International Nuclear Information System (INIS)

1975-01-01

A description is given of a cooling water intake collector for a nuclear reactor. It includes multiple sub-collectors extending out in a generally parallel manner to each other, each one having a first end and a second one separated along their length, and multiple water outlets for connecting each one to a corresponding pressure tube of the reactor. A first end tube and a second one connect the sub-collector tubes together to their first and second ends respectively. It also includes multiple collector tubes extending transversely by crossing over the sub-collector tubes and separated from each other in the direction of these tubes. Each collector tubes has a water intake for connecting to a water pump and multiple connecting tubes separated over its length and connecting each one to the corresponding sub-collector [fr

Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

International Nuclear Information System (INIS)

Carter, W.L.; Olsen, A.R.

1979-06-01

The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing

Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

Energy Technology Data Exchange (ETDEWEB)

Carter, W.L.; Olsen, A.R.

1979-06-01

The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing.

An improved water cooled nuclear reactor and pressuriser assembly

International Nuclear Information System (INIS)

Gardner, F.J.; Strong, R.

1991-01-01

A water cooled nuclear reactor is described which comprises a reactor core, a primary water coolant circuit and a pressuriser arranged as an integral unit in a pressure vessel. The pressure vessel is divided into an upper and a lower chamber by a casing. The reactor core and primary water coolant circuit are arranged in the lower chamber and the pressuriser is arranged in the upper chamber. A plurality of spray pipes interconnect a steam space of the pressuriser with the downcomer of the primary water coolant circuit below a heat exchanger. A plurality of surge ports interconnect a water space of the pressuriser with the primary water coolant circuit. The surge ports have hydraulic diodes so that there is a low flow resistance for water from the water space of the pressuriser to the primary water coolant circuit and high flow resistance in the opposite direction. The spray pipes provide a desuperheating spray of cooled water into the pressuriser during positive volume surges of the primary water coolant. The pressuriser arrangement may also be applied to integral water cooled reactors with separate pressurisers and to dispersed pressurised water reactors. The surge ports also allow water to flow by gravity to the core in an emergency. (author)

Balancing passive and active systems for evolutionary water cooled reactors

International Nuclear Information System (INIS)

Fil, N.S.; Allen, P.J.; Kirmse, R.E.; Kurihara, M.; Oh, S.J.; Sinha, R.K.

1999-01-01

Advanced concepts of the water-cooled reactors are intended to improve safety, economics and public perception of nuclear power. The potential inclusion of new passive means in addition or instead of traditional active systems is being considered by nuclear plant designers to reach these goals. With respect to plant safety, application of the passive means is mainly intended to simplify the safety systems and to improve their reliability, to mitigate the effect of human errors and equipment malfunction. However, some clear drawbacks and the limited experience and testing of passive systems may raise additional questions that have to be addressed in the design process for each advanced reactor. Therefore the plant designer should find a reasonable balance of active and passive means to effectively use their advantages and compensate their drawbacks. Some considerations that have to be taken into account when balancing active/passive means in advanced water-cooled reactors are discussed in this paper. (author)

Advanced applications of water cooled nuclear power plants

International Nuclear Information System (INIS)

2008-07-01

By August 2007, there were 438 nuclear power plants (NPPs) in operation worldwide, with a total capacity of 371.7 GW(e). Further, 31 units, totaling 24.1 GW(e), were under construction. During 2006 nuclear power produced 2659.7 billion kWh of electricity, which was 15.2% of the world's total. The vast majority of these plants use water-cooled reactors. Based on information provided by its Member States, the IAEA projects that nuclear power will grow significantly, producing between 2760 and 2810 billion kWh annually by 2010, between 3120 and 3840 billion kWh annually by 2020, and between 3325 and 5040 billion kWh annually by 2030. There are several reasons for these rising expectations for nuclear power: - Nuclear power's lengthening experience and good performance: The industry now has more than 12 000 reactor years of experience, and the global average nuclear plant availability during 2006 reached 83%; - Growing energy needs: All forecasts project increases in world energy demand, especially as population and economic productivity grow. The strategies are country dependent, but usually involve a mix of energy sources; - Interest in advanced applications of nuclear energy, such as seawater desalination, steam for heavy oil recovery and heat and electricity for hydrogen production; - Environmental concerns and constraints: The Kyoto Protocol has been in force since February 2005, and for many countries (most OECD countries, the Russian Federation, the Baltics and some countries of the Former Soviet Union and Eastern Europe) greenhouse gas emission limits are imposed; - Security of energy supply is a national priority in essentially every country; and - Nuclear power is economically competitive and provides stability of electricity price. In the near term most new nuclear plants will be evolutionary water cooled reactors (Light Water Reactors (LWRs) and Heavy Water Reactors (HWRs), often pursuing economies of scale. In the longer term, innovative designs that

Evolutionary water cooled reactors: Strategic issues, technologies and economic viability. Proceedings of a symposium

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-12-01

Symposium on evolutionary water cooled reactors: Strategic issues, technologies and economic viability was intended for managers in utilities, reactor design organizations and hardware manufacturing companies and for government decision makers who need to understand technological advances and the potential of evolutionary water cooled reactors to contribute to near and medium term energy demands. The topics addressed include: strategic issues (global energy outlook, the role of nuclear power in sustainable energy strategies, power generation costs, financing of nuclear plant projects, socio-political factors and nuclear safety requirements); technological advances (instrumentation and control, means od improving prevention and mitigation of severe accidents, development of passive safety systems); keys to economic viability (simplification, standardization, advances in construction and project management, feedback of experience from utilities into new designs, and effective management of plant operation)

Gas-cooled reactor power systems for space

International Nuclear Information System (INIS)

Walter, C.E.

1987-01-01

Efficiency and mass characteristics for four gas-cooled reactor power system configurations in the 2- to 20-MWe power range are modeled. The configurations use direct and indirect Brayton cycles with and without regeneration in the power conversion loop. The prismatic ceramic core of the reactor consists of several thousand pencil-shaped tubes made from a homogeneous mixture of moderator and fuel. The heat rejection system is found to be the major contributor to system mass, particularly at high power levels. A direct, regenerated Brayton cycle with helium working fluid permits high efficiency and low specific mass for a 10-MWe system

IAEA'S study on advanced applications of water cooled nuclear power plants

International Nuclear Information System (INIS)

Cleveland, J.; McDonald, A.; Rao, A.; )

2008-01-01

About one-fifth of the world's energy consumption is used for electricity generation, with nuclear power contributing approximately 15.2% of this electricity. However; most of the world's energy consumption is for heat and transportation. Nuclear energy has considerable potential to penetrate these energy sectors now served by fossil fuels that are characterized by price volatility and finite supply. Advanced applications of nuclear energy include seawater desalination, district heating, and heat for industrial processes. Nuclear energy also has potential to provide a near-term, greenhouse gas free, source of energy for transportation. These applications rely on a source of heat and electricity. Nuclear energy from water-cooled reactors, of course, is not unique in this sense. Indeed, higher temperature heat can be produced by burning natural gas and coal, or through the use of other nuclear technologies such as gas-cooled or liquid-metal-cooled reactors. Water-cooled reactors, however; are being deployed today while other reactor types have had considerably less operational and regulatory experience and will take still some time to be widely accepted in the market. Both seawater desalination and district heating with nuclear energy are well proven, and new seawater desalination projects using water-cooled reactors will soon be commissioned. Provision of process heat with nuclear energy can result in less dependence on fossil fuels and contribute to reductions of greenhouse gases. Importantly, because nuclear power produces base-load electricity at stable and predictable prices, it provides a greenhouse gas free source of electricity for transportation systems (trains and subways), and for electric and plug-in hybrid vehicles, and in the longer term nuclear energy could produce hydrogen for fuel cell vehicles, as well as for other components of a hydrogen economy. These advanced applications can play an important role in enhancing public acceptance of nuclear

Reactor core of light water-cooled reactor

International Nuclear Information System (INIS)

Miwa, Jun-ichi; Aoyama, Motoo; Mochida, Takaaki.

1996-01-01

In a reactor core of a light water cooled reactor, the center of the fuel rods or moderating rods situated at the outermost circumference among control rods or moderating rods are connected to divide a lattice region into an inner fuel region and an outer moderator region. In this case, the area ratio of the moderating region to the fuel region is determined to greater than 0.81 for every cross section of the fuel region. The moderating region at the outer side is increased relative to the fuel rod region at the inner side while keeping the lattice pitch of the fuel assembly constant, thereby suppressing the increase of an absolute value of a void reactivity coefficient which tends to be caused when using MOX fuels as a fuel material, by utilizing neutron moderation due to a large quantity of coolants at the outer side of the fuel region. The void reactivity coefficient can be made substantially equal with that of uranium fuel assembly without greatly reducing a plutonium loading amount or without greatly increasing linear power density. (N.H.)

High power density reactors based on direct cooled particle beds

Science.gov (United States)

Powell, J. R.; Horn, F. L.

Reactors based on direct cooled High Temperature Gas Cooled Reactor (HTGR) type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out along the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBRs) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed.

Once-through cycle, supercritical-pressure light water cooled reactor concept

Energy Technology Data Exchange (ETDEWEB)

Oka, Y.; Koshizuka, S. [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab

2001-07-01

Concept of once-through cycle, supercritical-pressure light water cooled reactors was developed. The research covered major aspects of conceptual design such as cores of thermal and fast reactors, plant system and heat balance, safety system and criteria, accident and transient analysis, LOCA, PSA, plant control and start-up. The advantages of the reactor lie in the compactness of the plant from high specific enthalpy of supercritical water, the simplicity of the once-through cycle and the experiences of major component technologies which are based on supercritical fossil-fired power plants and LWRs. The operating temperatures of the major components are within the experience in spite of high coolant outlet temperature. The once-through cycle is compatible with the tight fuel lattice fast reactor because of high head pumps and small coolant flow rate. (author)

Once-through cycle, supercritical-pressure light water cooled reactor concept

International Nuclear Information System (INIS)

Oka, Y.; Koshizuka, S.

2001-01-01

Concept of once-through cycle, supercritical-pressure light water cooled reactors was developed. The research covered major aspects of conceptual design such as cores of thermal and fast reactors, plant system and heat balance, safety system and criteria, accident and transient analysis, LOCA, PSA, plant control and start-up. The advantages of the reactor lie in the compactness of the plant from high specific enthalpy of supercritical water, the simplicity of the once-through cycle and the experiences of major component technologies which are based on supercritical fossil-fired power plants and LWRs. The operating temperatures of the major components are within the experience in spite of high coolant outlet temperature. The once-through cycle is compatible with the tight fuel lattice fast reactor because of high head pumps and small coolant flow rate. (author)

Specific power of liquid-metal-cooled reactors

International Nuclear Information System (INIS)

Dobranich, D.

1987-10-01

Calculations of the core specific power for conceptual space-based liquid-metal-cooled reactors, based on heat transfer considerations, are presented for three different fuel types: (1) pin-type fuel; (2) cermet fuel; and (3) thermionic fuel. The calculations are based on simple models and are intended to provide preliminary comparative results. The specific power is of interest because it is a measure of the core mass required to produce a given amount of power. Potential problems concerning zero-g critical heat flux and loss-of-coolant accidents are also discussed because these concerns may limit the core specific power. Insufficient experimental data exists to accurately determine the critical heat flux of liquid-metal-cooled reactors in space; however, preliminary calculations indicate that it may be a concern. Results also indicate that the specific power of the pin-type fuels can be increased significantly if the gap between the fuel and the clad is eliminated. Cermet reactors offer the highest specific power because of the excellent thermal conductivity of the core matrix material. However, it may not be possible to take fuel advantage of this characteristic when loss-of-coolant accidents are considered in the final core design. The specific power of the thermionic fuels is dependent mainly on the emitter temperature. The small diameter thermionic fuels have specific powers comparable to those of pin-type fuels. 11 refs., 12 figs, 2 tabs

Method of 16N generation for test of radiation controlled channels at nuclear power stations with water-cooled reactors

International Nuclear Information System (INIS)

Khryachkov, V.A.; Bondarenko, I.P.; Dvornikov, P.A.; Zhuravlev, B.V.; Kovtun, S.N.; Khromyleva, T.A.; Pavlov, A.V.; Roshchin, N.G.

2012-01-01

The preferences of nuclear reaction use for radiation control channels test in water-cooled power reactors have been analyzed in the paper. The new measurements for more accurate determination of reaction cross section energy dependence have been carried out. A set of new methods for background reducing and improvement of events determination reliability has also been developed [ru

Hydrogen behaviour and mitigation in water-cooled nuclear power reactors

International Nuclear Information System (INIS)

Della Loggia, E.

1992-01-01

The Commission of the European Communities (CEC) and the International Atomic Energy Agency (IAEA), within the framework of their safety research activities, initiated and arranged a series of specialist meetings and research contracts on hydrogen behaviour and control. The result of this work is summarized in a report jointly prepared by the two international organizations entitled 'Hydrogen in water-cooled nuclear power reactors'. Independently, the Kurchatov Atomic Energy Institute organized a workshop on the hydrogen issue in Sukhumi, USSR, with CEC and IAEA cooperation. Commonly expressed views have emerged and recommendations were formulated to organize the subsequent seminar/workshop concentrating mainly on the most recent research and analytical projects and findings related to the hydrogen behaviour, and-most importantly-on the practical approaches and engineering solutions to the hydrogen control and mitigation. The seminar/workshop, therefore, addressed the 'theory and practice' aspects of the hydrogen issue. The workshop was structured in the following sessions: combustible gas production; hydrogen distribution; combustion phenomena; combustion effects and threats; and detection and migration

Solid-Core, Gas-Cooled Reactor for Space and Surface Power

International Nuclear Information System (INIS)

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

2006-01-01

The solid-core, gas-cooled, Submersion-Subcritical Safe Space (S and 4) reactor is developed for future space power applications and avoidance of single point failures. The Mo-14%Re reactor core is loaded with uranium nitride fuel in enclosed cavities, cooled by He-30%Xe, and sized to provide 550 kWth for seven years of equivalent full power operation. The beryllium oxide reflector disassembles upon impact on water or soil. In addition to decreasing the reactor and shadow shield mass, Spectral Shift Absorber (SSA) materials added to the reactor core ensure that it remains subcritical in the worst-case submersion accident. With a 0.1 mm thick boron carbide coating on the outside surface of the core block and 0.25 mm thick iridium sleeves around the fuel stacks, the reflector outer diameter is 43.5 cm and the combined reactor and shadow shield mass is 935.1 kg. With 12.5 atom% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide intersititial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating, the S and 4 reactor has a slightly smaller reflector outer diameter of 43.0 cm, and a total reactor and shield mass of 901.7 kg. With 8.0 atom% europium-151 added to the fuel, 2.0 mm diameter europium-151 sesquioxide interstitial pins, and a 0.1 mm thick europium-151 sesquioxide 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

Device for preventing cooling water from flowing out of reactor

International Nuclear Information System (INIS)

Chinen, Masanori; Kotani, Koichi; Murase, Michio.

1976-01-01

Object: To provide emergency cooling system, which can prevent cooling water bearing radioactivity from flowing to the outside of the reactor at the time of breakage of feedwater pipe, thus eliminating the possibility of exposure of the fuel rod to provide high reliability and also reducing the possibility of causing radioactive pollution. Structure: The device for preventing cooling water from flowing out from the reactor features a jet nozzle inserted in a feedwater pipe adjacent to the inlet or outlet thereof immediately before the reactor container. The nozzle outlet is provided in the vicinity of the reactor wall and in a direction opposite to the direction of out-flow, and water supplied from a high pressure pump is jetted from it. (Nakamura, S.)

Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

Science.gov (United States)

Kalyakin, S. G.; Kirillov, P. L.; Baranaev, Yu. D.; Glebov, A. P.; Bogoslovskaya, G. P.; Nikitenko, M. P.; Makhin, V. M.; Churkin, A. N.

2014-08-01

The state of nuclear power engineering as of February 1, 2014 and the accomplished elaborations of a supercritical-pressure water-cooled reactor are briefly reviewed, and the prospects of this new project are discussed based on this review. The new project rests on the experience gained from the development and operation of stationary water-cooled reactor plants, including VVERs, PWRs, BWRs, and RBMKs (their combined service life totals more than 15 000 reactor-years), and long-term experience gained around the world with operation of thermal power plants the turbines of which are driven by steam with supercritical and ultrasupercritical parameters. The advantages of such reactor are pointed out together with the scientific-technical problems that need to be solved during further development of such installations. The knowledge gained for the last decade makes it possible to refine the concept and to commence the work on designing an experimental small-capacity reactor.

Cooling water requirements and nuclear power plants

International Nuclear Information System (INIS)

Rao, T.S.

2010-01-01

Indian nuclear power programme is poised to scuttle the energy crisis of our time by proposing joint ventures for large power plants. Large fossil/nuclear power plants (NPPs) rely upon water for cooling and are therefore located near coastal areas. The amount of water a power station uses and consumes depends on the cooling technology used. Depending on the cooling technology utilized, per megawatt existing NPPs use and consume more water (by a factor of 1.25) than power stations using other fuel sources. In this context the distinction between 'use' and 'consume' of water is important. All power stations do consume some of the water they use; this is generally lost as evaporation. Cooling systems are basically of two types; Closed cycle and Once-through, of the two systems, the closed cycle uses about 2-3% of the water volumes used by the once-through system. Generally, water used for power plant cooling is chemically altered for purposes of extending the useful life of equipment and to ensure efficient operation. The used chemicals effluent will be added to the cooling water discharge. Thus water quality impacts on power plants vary significantly, from one electricity generating technology to another. In light of massive expansion of nuclear power programme there is a need to develop new ecofriendly cooling water technologies. Seawater cooling towers (SCT) could be a viable option for power plants. SCTs can be utilized with the proper selection of materials, coatings and can achieve long service life. Among the concerns raised about the development of a nuclear power industry, the amount of water consumed by nuclear power plants compared with other power stations is of relevance in light of the warming surface seawater temperatures. A 1000 MW power plant uses per day ∼800 ML/MW in once through cooling system; while SCT use 27 ML/MW. With the advent of new marine materials and concrete compositions SCT can be constructed for efficient operation. However, the

Cooling Performance of Natural Circulation for a Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Park, Suki; Chun, J. H.; Yum, S. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

This paper deals with the core cooling performance by natural circulation during normal operation and a flow channel blockage event in an open tank-in-pool type research reactor. The cooling performance is predicted by using the RELAP5/ MOD3.3 code. The core decay heat is usually removed by natural circulation to the reactor pool water in open tank-in-pool type research reactors with the thermal power less than several megawatts. Therefore, these reactors have generally no active core cooling system against a loss of normal forced flow. In reactors with the thermal power less than around one megawatt, the reactor core can be cooled down by natural circulation even during normal full power operation. The cooling performance of natural circulation in an open tank-in-pool type research reactor has been investigated during the normal natural circulation and a flow channel blockage event. It is found that the maximum powers without void generation at the hot channel are around 1.16 MW and 820 kW, respectively, for the normal natural circulation and the flow channel blockage event.

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

Science.gov (United States)

Corletti, Michael M.; Lau, Louis K.; Schulz, Terry L.

1993-01-01

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

Elements of Design Consideration of Once-Through Cycle, Supercritical-Pressure Light Water Cooled Reactor

International Nuclear Information System (INIS)

Yoshiaki Oka; Sei-ichi Koshizuka; Yuki Ishiwatari; Akifumi Yamaji

2002-01-01

The paper describes elements of design consideration of supercritical-pressure, light water cooled reactors as well as the status and prospects of the research and development. It summarizes the results of the conceptual design study at the University of Tokyo from 1989. The research and development started in Japan, Europe and USA. The major advantages of the reactors are 1. Compact reactor and turbines due to high specific enthalpy of supercritical water 2.Simple plant system because of the once-through coolant cycle 3.Use of the experience of LWR and fossil-fired power plants. The temperatures of the major components such as reactor pressure vessel, coolant pipes, pumps and turbines are within the experience, in spite of the high outlet coolant temperature. 4.Similarity to LWR safety design and criteria, but no burnout phenomenon 5.Potential cost reduction due to smaller material expenditure and short construction period 6.The smallest reactor not in power rating, but in plant sizes. 7.High-thermal efficiency and low coolant flow rate because of high enthalpy rise. 8.Water cooled reactors potentially free from SCC (stress corrosion cracking) problems. 9.Compatibility of tight-fuel-lattice fast reactor core due to small coolant flow rate, potentially easy shift to fast breeder reactor without changing coolant technology. 10.Potential of producing energy products such as hydrogen and high quality hydro carbons. (authors)

Advanced light-water reactors

International Nuclear Information System (INIS)

Golay, M.W.; Todreas, N.E.

1990-01-01

Environmental concerns, economics and the earth's finite store of fossil fuels argue for a resuscitation of nuclear power. The authors think improved light-water reactors incorporating passive safety features can be both safe and profitable, but only if attention is paid to economics, effective management and rigorous training methods. The experience of nearly four decades has winnowed out designs for four basic types of reactor: the heavy-water reactor (HWR), the gas-cooled rector (GCR), the liquid-metal-cooled reactor (LMR) and the light-water reactor (LWR). Each design is briefly described before the paper discusses the passive safety features of the AP-600 rector, so-called because it employs an advanced pressurized water design and generates 600 MW of power

Water chemistry in nuclear power stations with high-temperature reactors with particular reference to the AVR

International Nuclear Information System (INIS)

Nieder, R.; Resch, G.

1976-01-01

The water-steam cycle of a nuclear power plant with a helium-cooled high-temperature reactor differs in design data significantly and extensively from the corresponding cycles of light-water-cooled nuclear reactors and resembles to a great extent the water-steamcycle of a modern conventional power plant. The radioactive constituents of the water-steam cycle can be satisfactorily removed apart from Tritium by means of a pre-coat filter with powder-resisn, as comprehensive experiments have demonstrated. (orig.) [de

Thermophysical properties of materials for water cooled reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-06-01

The IAEA Co-ordinated Research Programme (CRP) to establish a thermophysical properties data base for light and heavy water reactor materials was organized within the framework of the IAEA`s International Working Group on Advanced Technologies for Water Cooled Reactors. The work within the CRP started in 1990. The objective of the CRP was to collect and systemaize a thermophysical properties data base for light and heavy water reactor materials under normal operating, transient and accident conditions. The important thermophysical properties include thermal conductivity, thermal diffusivity, specific heat capacity, enthalpy, thermal expansion and others. These properties as well as the oxidation of zirconium-based alloys, the thermophysical characteristics of high temperature concrete-core melt interaction and the mechanical properties of construction materials are presented in this report. It is hoped that this report will serve as a useful source of thermophysical properties data for water cooled reactor analyses. The properties data are maintained on the THERSYST system at the University of Stuttgart, Germany and are internationally available. Refs, figs, tabs.

Thermophysical properties of materials for water cooled reactors

International Nuclear Information System (INIS)

1997-06-01

The IAEA Co-ordinated Research Programme (CRP) to establish a thermophysical properties data base for light and heavy water reactor materials was organized within the framework of the IAEA's International Working Group on Advanced Technologies for Water Cooled Reactors. The work within the CRP started in 1990. The objective of the CRP was to collect and systemaize a thermophysical properties data base for light and heavy water reactor materials under normal operating, transient and accident conditions. The important thermophysical properties include thermal conductivity, thermal diffusivity, specific heat capacity, enthalpy, thermal expansion and others. These properties as well as the oxidation of zirconium-based alloys, the thermophysical characteristics of high temperature concrete-core melt interaction and the mechanical properties of construction materials are presented in this report. It is hoped that this report will serve as a useful source of thermophysical properties data for water cooled reactor analyses. The properties data are maintained on the THERSYST system at the University of Stuttgart, Germany and are internationally available. Refs, figs, tabs

Thermal calculations for water cooled research reactors

International Nuclear Information System (INIS)

Fabrega, S.

1979-01-01

The formulae and the more important numerical data necessary for thermic calculations on the core of a research reactor, cooled with low pressure water, are presented. Most of the problems met by the designer and the operator are dealt with (calculations margins, cooling after shut-down). Particular cases are considered (gas release, rough walls, asymmetric cooling slabs etc.), which are not generally envisaged in works on general thermics

Technological readiness of evolutionary water cooled reactors

International Nuclear Information System (INIS)

Juhn, P.E.

1999-01-01

Nuclear energy has evolved to a mature industry that supplies over 16% of the world's electricity, and it represents an important option for meeting the global energy demands of the coming century in an environmentally acceptable manner. New, evolutionary water cooled reactor designs that build on successful performance of predecessors have been developed; these designs have generally been guided by wishes to reduce cost, to improve availability and reliability, and to meet increasingly stringent safety objectives. These three aspects are important factors in what has been called technological readiness for an expanded deployment of nuclear power; a major increase in utilization of nuclear power will only occur if it is economically competitive, and meets safety expectations. To this end, the industry will also have to maintain or improve the public perception of nuclear power as a benign, economical and reliable energy source. (author)

The early history of high-temperature helium gas-cooled nuclear power reactors

International Nuclear Information System (INIS)

Simnad, M.T.; California Univ., San Diego, La Jolla, CA

1991-01-01

The original concepts in the proposals for high-temperature helium gas-cooled power reactors by Farrington Daniels, during the decade 1944-1955, are summarized. The early research on the development of the helium gas-cooled power reactors is reviewed, and the operational experiences with the first generation of HTGRs are discussed. (author)

Gas-cooled reactors

International Nuclear Information System (INIS)

Schulten, R.; Trauger, D.B.

1976-01-01

Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing

Method of inhibiting concentration of radioactive corrosion products in cooling water or nuclear power plants

International Nuclear Information System (INIS)

Takabayashi, Jun-ichi; Hishida, Mamoru; Ishikura, Takeshi.

1979-01-01

Purpose: To suppress the increase in the concentration of the radioactive corrosion products in cooling water, which increase is accompanied by the transference of the corrosion products activated and accumulated in the core due to dissolution and exfoliation into the core water, and inhibit the flowing of said products out of the core and the diffusion thereof into the cooling system, thereby to prevent the accumulation of said products in the cooling system and prevent radioactive contaminations. Method: In a nuclear power plant of a BWR type light water reactor, when the temperature of the pile water is t 0 C, hydrogen is injected in cooling water in a period of time from immediately before starting of the drive stopping operation of the nuclear power plant to immediately after the termination of restarting operation, whereby the concentration of hydrogen in the reactor water through said period is maintained at a value more than 2exp (0.013 t) cm 3 N.T.P./kg H 2 O. (Aizawa, K.)

Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-03-01

This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

Passive cooling systems in power reactors

International Nuclear Information System (INIS)

Aharon, J.; Harrari, R.; Weiss, Y.; Barnea, Y.; Katz, M.; Szanto, M.

1996-01-01

This paper reviews several R and D activities associated with the subject of passive cooling systems, conducted by the N.R.C.Negev thermohydraulic group. A short introduction considering different types of thermosyphons and their applications is followed by a detailed description of the experimental work, its results and conclusions. An ongoing research project is focused on the evaluation of the external dry air passive containment cooling system (PCCS) in the AP-600 (Westinghouse advanced pressurized water reactor). In this context some preliminary theoretical results and planned experimental research are for the fature described

Uranium utilization of light water cooled reactors and fast breeders

International Nuclear Information System (INIS)

Stojadinovic, Timm

1991-08-01

The better uranium utilization of fast breeder reactors as compared with water cooled reactors is one argument in favour of the breeder introduction. This report tries to quantify this difference. It gives a generally valid formalism for the uranium utilization as a function of the fuel burnup, the conversion rate, fuel cycle losses and the fuel enrichment. On the basis of realistic assumptions, the ratio between the utilizations of breeder reactors to that of light water cooled reactors (LWR) amounts to 180 for the open LWR cycle and 100 in case of plutonium recycling in LWRs

Emergency reactor core cooling facility

International Nuclear Information System (INIS)

Yoshikawa, Kazuhiro; Kinoshita, Shoichiro; Iwata, Yasutaka.

1996-01-01

The present invention provides an emergency reactor core cooling device for a BWR type nuclear power plant. Namely, D/S pit (gas/water separator storage pool) water is used as a water source for the emergency reactor core cooling facility upon occurrence of loss of coolant accidents (LOCA) by introducing the D/S pit water to the emergency reactor core cooling (ECCS) pump. As a result, the function as the ECCS facility can be eliminated from the function of the condensate storage tank which has been used as the ECCS facility. If the function is unnecessary, the level of quality control and that of earthquake resistance of the condensate storage tank can be lowered to a level of ordinary facilities to provide an effect of reducing the cost. On the other hand, since the D/S pit as the alternative water source is usually a facility at high quality control level and earthquake resistant level, there is no problem. The quality of the water in the D/S pit can be maintained constant by elevating pressure of the D/S pit water by a suppression pool cleanup (SPCU) pump to pass it through a filtration desalter thereby purifying the D/S pit water during the plant operation. (I.S.)

Emergency reactor core cooling facility

Energy Technology Data Exchange (ETDEWEB)

Yoshikawa, Kazuhiro; Kinoshita, Shoichiro; Iwata, Yasutaka

1996-11-01

The present invention provides an emergency reactor core cooling device for a BWR type nuclear power plant. Namely, D/S pit (gas/water separator storage pool) water is used as a water source for the emergency reactor core cooling facility upon occurrence of loss of coolant accidents (LOCA) by introducing the D/S pit water to the emergency reactor core cooling (ECCS) pump. As a result, the function as the ECCS facility can be eliminated from the function of the condensate storage tank which has been used as the ECCS facility. If the function is unnecessary, the level of quality control and that of earthquake resistance of the condensate storage tank can be lowered to a level of ordinary facilities to provide an effect of reducing the cost. On the other hand, since the D/S pit as the alternative water source is usually a facility at high quality control level and earthquake resistant level, there is no problem. The quality of the water in the D/S pit can be maintained constant by elevating pressure of the D/S pit water by a suppression pool cleanup (SPCU) pump to pass it through a filtration desalter thereby purifying the D/S pit water during the plant operation. (I.S.)

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Science.gov (United States)

2010-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

Vaporization Rate Analysis of Primary Cooling Water from Reactor PUSPATI TRIGA (RTP) Tank

International Nuclear Information System (INIS)

Tonny Anak Lanyau; Mohd Fazli Zakaria; Yahya Ismail

2011-01-01

Primary cooling system consists of pumps, heat exchangers, probes, a nitrogen-16 diffuser and associated valves is connected to the reactor TRIGA PUSPATI (RTP) tank by aluminium pipes. Both the primary cooling system and the reactor tank is filled with demineralized light water (H 2 O), which serves as a coolant, moderator as well as shielding. During reactor operation, vaporization in the reactor tank will reduce the primary water and contribute to the formation of vapor in the reactor hall. The vaporization may influence the function of the water subsequently may affect the safety of the reactor operation. It is essential to know the vaporization rate of the primary water to ensure its functionality. This paper will present the vaporization rate of the primary cooling water from the reactor tank and the influence of temperature of the water in the reactor tank to the vaporization rate. (author)

Power maximization method for land-transportable fully passive lead–bismuth cooled small modular reactor systems

Energy Technology Data Exchange (ETDEWEB)

Cho, Jaehyun, E-mail: chojh@kaeri.re.kr [Korea Atomic Energy Research Institute, 1405 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Shin, Yong-Hoon; Hwang, Il Soon [Seoul National University, Sillim-dong, Gwanak-gu, Seoul 151-742 (Korea, Republic of)

2015-08-15

Highlights: • The power maximization method for LBE natural circulation cooled SMRs was developed. • The two powers in view of neutronics and thermal-hydraulics were considered. • The limitations for designing of LBE natural circulation cooled SMRs were summarized. • The necessary conditions for safety shutdown in accidents were developed. • The maximized power in the case study is 206 MW thermal. - Abstract: Although current pressurized water reactors (PWRs) have significantly contributed to global energy supply, PWR technology has not been considered a trustworthy energy solution owing to its problems of spent nuclear fuels (SNFs), nuclear safety, and nuclear economy. In order to overcome these problems, a lead–bismuth eutectic (LBE) fully passive cooling small modular reactor (SMR) system is suggested. This technology can not only provide the solution for the problems of SNFs through the transmutation feature of the LBE coolant, but also strengthen safety and economy through the concept of natural circulation cooling SMRs. It is necessary to maximize the advantages, namely safety and economy, of this type of nuclear power plants for broader applications in the future. Accordingly, the objective of this study is to maximize the reactor core power while satisfying the limitations of shipping size, materials endurance, and criticality of a long-burning core as well as safety under beyond design basis events. To achieve these objectives, the design limitations of natural circulating LBE-cooling SMRs are derived. Then, the power maximization method is developed based on obtaining the design limitations. The results of this study are expected to contribute to the effectiveness of the reactor design stage by providing insights to designers, as well as by formulating methods for the power maximization of other types of SMRs.

The role of the IAEA in advanced technologies for water-cooled reactors

International Nuclear Information System (INIS)

Cleveland, J.

1996-01-01

The role of the IAEA in advanced technologies for water-cooled reactors is described, including the following issues: international collaboration ways through international working group activities; IAEA coordinated research programmes; cooperative research in advanced water-cooled reactor technology

Steam generator tube performance. Experience with water-cooled nuclear power reactors during 1985

International Nuclear Information System (INIS)

Tatone, O.S.; Tapping, R.L.

1988-12-01

The performance of steam generator tubes at water-cooled reactors during 1985 has been reviewed. Seventy-three of 168 reactors in the survey experienced tube degradation sufficient for the tubes to be plugged. The number of tubes plugged was 6837 or 0.28% of those in service. The leading cause of tube failure was stress corrosion cracking from the primary side. Stress corrosion cracking or intergranular attack from the secondary side and pitting were also major causes of tube failure. Unlike most previous years, fretting was a substantial problem at some reactors. Overall, corrosion continued to account for more than 80% of the defects. 20 refs

Implementation of new core cooling monitoring system for light water reactors - BCCM (Becker Core Cooling Monitor)

International Nuclear Information System (INIS)

Coville, Patrick; Eliasson, Bengt; Stromqvist, Erik; Ward, Olav; Fox, Georges; Ashjian, D. T.

1998-01-01

Core cooling monitors are key instruments to protect reactors from large accidents due to loss of coolant. Sensors presented here are based on resistance thermometry. Temperature dependent resistance is powered by relatively high and constant current. Value of this resistance depends on thermal exchange with coolant and when water is no more surrounding the sensors a large increase of temperature is immediately generated. The same instrument can be operated with low current and will measure the local temperature up to 1260 o C in case of loss of coolant accident. Sensors are manufactured with very few components and materials already qualified for long term exposure to boiling or pressurized water reactors environment. Prototypes have been evaluated in a test loop up to 160 bars and in the Barsebaeck-1 reactor. Industrial sensors are now in operation in reactor Oskarshamn 2. (author)

In-service inspections of the reactor cooling system of pressurized water reactors

International Nuclear Information System (INIS)

Fuerste, W.; Hohnerlein, G.; Werden, B.

1982-01-01

In order to guarantee constant safety of the components of the reactor cooling system, regular in-service inspections are carried out after commissioning of the nuclear power plant. This contribution is concerned with the components of the reactor cooling system, referring to the legal requirements, safety-related purposes and scope of the in-service inspections during the entire period of operation of a nuclear power plant. Reports are made with respect to type, examination intervals, examination technique, results and future development. The functional tests which are carried out within the scope of the in-service inspections are not part of this contribution. (orig.) [de

Gas-cooled reactor for space power systems

International Nuclear Information System (INIS)

Walter, C.E.; Pearson, J.S.

1987-05-01

Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors

Surveillance tests for light-water cooled nuclear power reactor vessels in IMEF

International Nuclear Information System (INIS)

Choo, Yong-Sun; Ahn, Sang-Bok; Park, Dae-Gyu; Jung, Yang-Hong; Yoo, Byung-Ok; Oh, Wan-Ho; Baik, Seung-Je; Koo, Dae-Seo; Lee, Key-Soon

1999-01-01

The surveillance tests for light-water cooled nuclear power reactor vessels were established to monitor the radiation-induced changes in the mechanical properties of ferritic materials in the beltline according to US NRC 10 CFR 50 App. G, US NRC RG1.99-rev.2, ASTM E185-82 and E185-94 in Irradiated Materials Examination Facility(IMEF). The surveillance capsule was transported from NPPs pool sites to KAERI IMEF by using a shipping cask. The capsule was cut and dismantled by capsule cutting machine and milling machine in M2 hot cell. Charpy tests and tension tests were performed in M5a and M5b hot cells respectively. Especially the EPMA located at hot lab was used to analyze the Ni and Cu wt% composition of base metal and weld for predicting the adjusted reference temperature(ART). The established process and test results were summarized in this paper. (author)

Proceedings (slides) of the OECD/NEA Workshop on Innovations in Water-cooled Reactor Technologies

International Nuclear Information System (INIS)

Spiler, Joze; Kim, Sang-Baik; ); Feron, Fabien; Jaervinen, Marja-Leena; Husse, Julien; ); Ferraro, Giovanni; Bertels, Frank; Denk, Wolfgang; Tuomisto, Harri; Golay, Michael; Buongiorno, J.; Todreas, N.; Adams, E.; Briccetti, A.; Jurewicz, J.; Kindfuller, V.; Srinivasan, G.; Strother, M.; Minelli, P.; Fasil, E.; Zhang, J.; Genzman, G.; Epinois, Bertrand de l'; Kim, Shin Whan; Laaksonen, Jukka; Maltsev, Mikhail; Yu, CHongxing; Powell, David; Gorgemans, Julie; Hopwood, Jerry; Bylov, Igor; Bakhmetyev, Alexander M.; Lepekhin, Andrey N.; Fadeev, Yuriy P.; Bruna, Giovanni; Gulliford, Jim; ); Ham-Su, Rosaura; Thevenot, Caroline; GAUTIER, Guy-Marie; MARSAULT, Philippe; PIGNATEL, Jean-Francois; White, Andrew; )

2015-02-01

New technologies and solutions have been developed over more than thirty years to improve the safety, performance and economics of nuclear power plants. Particular efforts were made in designing systems to prevent or mitigate nuclear accidents and, greatly limit or even avoid any offsite release of radioactivity. Reactor designs developed in the 1980's and later are often referred to as Generation III (Gen III) reactors. They offer enhanced safety compared to earlier Generation II (Gen II) designs, as well as improved performance and economics. Examples of Gen III safety design features include solutions for corium localisation, advanced containment structures, improved emergency core-cooling systems, filtered venting systems, hydrogen risk management solutions, etc. Some of these solutions have also been back-fitted or partially adapted to existing reactors, based on recommendations from regulators or modernisation efforts by the utilities operating these reactors, to bring their level of safety to levels approaching those of the more modern designs. Other innovations found in the latest water-cooled reactor designs include the use of passive safety systems, and often associated with those, a simplification in the design of the reactor. Gen III reactors also feature better economics, for example increased design lifetime up to 60 years, ability to use 100% MOX fuel and operate with higher flexibility, higher thermal efficiencies and reduced staff requirements. Modularity is often quoted as a feature of some Gen III designs as a way of reducing the construction times and simplifying the decommissioning of the plant. The scope of the Workshop includes, inter alia: - Evolution of regulatory and design requirements for commercial water-cooled reactors; - Innovations in water-cooled reactor technologies that allowed significant improvement in the level of safety, with a discussion on advantages and challenges of active vs. passive safety systems; - Innovations under

Stability monitoring of a natural-circulation-cooled boiling water reactor

International Nuclear Information System (INIS)

Hagen, T.H.J.J. van der.

1989-01-01

Methods for monitoring the stability of a boiling water reactor (BWR) are discussed. Surveillance of BWR stability is of importance as problems were encountered in several large reactors. Moreover, surveying stability allows plant owners to operate at high power with acceptable stability margins. The results of experiments performed on the Dodewaard BWR (the Netherlands) are reported. This type reactor is cooled by natural circulation, a cooling principle that is also being considered for new reactor designs. The stability of this reactor was studied both with deterministic methods and by noise analysis. Three types of stability are distinguished and were investigated separately: reactor-kinetic stability, thermal-hydraulic stability and total-plant stability. It is shown that the Dodewaard reactor has very large stability margins. A simple yet reliable stability criterion is introduced. It can be derived on-line from thhe noise signal of ex-vessel neutron detectors during normal operation. The sensitivity of neutron detectors to in-core flux perturbations - reflected in the field-of-view of the detector - was calculated in order to insure proper stability surveillance. A novel technique is presented which enables the determination of variations of the in-core coolant velocity by noise correlation. The velocity measured was interpreted on the basis of experiments performed on the air/water flow in a model of a BWR coolant channel. It appeared from this analysis that the velocity measured was much higher than the volume-averaged water and air velocities and the volumetric flux. The applicability of the above-mentioned technique to monitoring of local channel-flow stability was tested. It was observed that stability effects on the coolant velocity are masked by other effects originating from the local flow pattern. Experimental and theoretical studies show a shorter effective fuel time constant in a BWR than was assumed. (author). 118 refs.; 73 figs.; 21 tabs

Design Requirements of an Advanced HANARO Reactor Core Cooling System

International Nuclear Information System (INIS)

Park, Yong Chul; Ryu, Jeong Soo

2007-12-01

An advanced HANARO Reactor (AHR) is an open-tank-type and generates thermal power of 20 MW and is under conceptual design phase for developing it. The thermal power is including a core fission heat, a temporary stored fuel heat in the pool, a pump heat and a neutron reflecting heat in the reflector vessel of the reactor. In order to remove the heat load, the reactor core cooling system is composed of a primary cooling system, a primary cooling water purification system and a reflector cooling system. The primary cooling system must remove the heat load including the core fission heat, the temporary stored fuel heat in the pool and the pump heat. The purification system must maintain the quality of the primary cooling water. And the reflector cooling system must remove the neutron reflecting heat in the reflector vessel of the reactor and maintain the quality of the reflector. In this study, the design requirement of each system has been carried out using a design methodology of the HANARO within a permissible range of safety. And those requirements are written by english intend to use design data for exporting the research reactor

Accident analysis of heavy water cooled thorium breeder reactor

International Nuclear Information System (INIS)

Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki

2015-01-01

Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

Accident analysis of heavy water cooled thorium breeder reactor

Energy Technology Data Exchange (ETDEWEB)

Yulianti, Yanti [Department of Physics, University of Lampung Jl. Sumantri Brojonegoro No.1 Bandar Lampung, Indonesia Email: y-yanti@unila.ac.id (Indonesia); Su’ud, Zaki [Department of Physics, Bandung Institute of Technology Jl. Ganesha 10 Bandung, Indonesia Email: szaki@fi.itb.ac.id (Indonesia); Takaki, Naoyuki [Department of Nuclear Safety Engineering Cooperative Major in Nuclear Energy (Graduate School) 1-28-1 Tamazutsumi,Setagayaku, Tokyo158-8557, Japan Email: ntakaki@tcu.ac.jp (Japan)

2015-04-16

Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

IAEA activities in technology development for advanced water-cooled nuclear power plants

International Nuclear Information System (INIS)

Juhn, Poong Eil; Kupitz, Juergen; Cleveland, John; Lyon, Robert; Park, Je Won

2003-01-01

As part of its Nuclear Power Programme, the IAEA conducts activities that support international information exchange, co-operative research and technology assessments and advancements with the goal of improving the reliability, safety and economics of advanced water-cooled nuclear power plants. These activities are conducted based on the advice, and with the support, of the IAEA Department of Nuclear Energy's Technical Working Groups on Advanced Technologies for Light Water Reactors (LWRs) and Heavy Water Reactors (HWRs). Assessments of projected electricity generation costs for new nuclear plants have shown that design organizations are challenged to develop advanced designs with lower capital costs and short construction times, and sizes, including not only large evolutionary plants but also small and medium size plants, appropriate to grid capacity and owner financial investment capability. To achieve competitive costs, both proven means and new approaches should be implemented. The IAEA conducts activities in technology development that support achievement of improved economics of water-cooled nuclear power plants (NPPs). These include fostering information sharing and cooperative research in thermo-hydraulics code validation; examination of natural circulation phenomena, modelling and the reliability of passive systems that utilize natural circulation; establishment of a thermo-physical properties data base; improved inspection and diagnostic techniques for pressure tubes of HWRs; and collection and balanced reporting from recent construction and commissioning experiences with evolutionary water-cooled NPPs. The IAEA also periodically publishes Status Reports on global development of advanced designs. (author)

Design measures in evolutionary water cooled reactors to optimize for economic viability

International Nuclear Information System (INIS)

Oh, S.J.; Yu, S.K.W.; Appell, B.

1999-01-01

Since the mid 1980s, there have been various efforts to develop evolutionary water cooled reactors based on the current operating plant experience. To sustain and improve the economic viability, particular attention has been paid to the following aspects in developing evolutionary water cooled reactors: design simplification and increased operating margins, standardization in design as well as construction and operation, integration of operating plant insights, and consideration of safety, operability and constructability during the design stage. This paper reviews each item and discusses several examples from some of the evolutionary water cooled reactors being developed. (author)

Research and development of the supercritical-pressure light water cooled reactor

International Nuclear Information System (INIS)

Oka, Yoshiaki

2003-01-01

The concept of high temperature reactor cooled by light water (SCR) has been developed at the University of Tokyo since 1989. Major elements of reactor conceptual design and safety were studied. It includes fuel rod design, core design of thermal and fast reactors, plant heat balance, safety design, accident and transient analysis, LOCA, PSA, plant control, start-up and stability. The big advantage of the SCR concept is that the temperatures of major components such as reactor pressure vessel, control rod drive mechanisms, containments, coolant pumps, main steam piping and turbines are within the temperatures of the components of LWR and supercritical FPP in spite of the high outlet coolant temperature. The experience of these components of LWR and supercritical fossil Fired Power Plants (FPP) will be fully utilized for SCR. Although the concept was developed at the University of Tokyo mostly with our own funds and resources, four funding was/is provided for the research in Japan so far. Those are TEPCO studies with Japanese vendors in 1994 and 1995. JSPS (Monbusho) funding of pulse radiolysis of supercritical water to the University of Tokyo, Japanese-NERI program of METI to Toshiba team on thermal hydraulics, corrosion and plant system and Japanese-NERI program of MEXT on water chemistry to the University of Tokyo. The concept was taken as the reference of HPLWR study in Europe with funding of EU in 2000 and 2001. The concept was evaluated in the Generation 4 reactor program in USA. It was selected as only one water-cooled Generation 4 reactor. This paper describes the overview of the conceptual design at the University of Tokyo and R and D in the world

Reactors of different types in the world nuclear power

International Nuclear Information System (INIS)

Simonov, K.V.

1991-01-01

The status of the world nuclear power is briefly reviewed. It is noted that PWR reactors have decisive significance in the world power. The second place is related to gas-cooled graphite-moderated reactors. Channel-type heavy water moderated reactors are relatively important. Nuclear power future is associated with fast liquid-metal cooled breeder reactors

Some methods of failed fuel element detection in water cooled reactors

International Nuclear Information System (INIS)

Strindehag, O.M.

1976-01-01

The methods are surveyed using fission products released in the coolant for the detection of failed fuel elements in water cooled reactors. The classification of the detection methods is made with respect to fission product detection in the coolant and to gaseous fission product detection. The detection systems are listed used for the AGESTA power reactor and for the experimental loops of the RA research reactor based on the detection of either gaseous fission products or gaseous daughter products. The AGESTA reactor detection systems using electrostatic precipitators consist of five precipitator channels of which three are intended for detection and two for localization. A special detection unit was developed for the failed fuel element detection in the R-2 reactor experimental steam loop. Its description is listed. In the reactor pressurized-water loop a Cherenkov counter was used in the detection of fission products. An ion exchange monitor whose application is described was used in the total measurement of the main coolant flow in the AGESTA reactor. (J.P.)

Cooling System Design Options for a Fusion Reactor

Science.gov (United States)

Natalizio, Antonio; Collén, Jan; Vieider, Gottfried

1997-06-01

The objective of a fusion power reactor is to produce electricity safely and reliably. Accordingly, the design, objective of the heat transport system is to optimize power production, safety, and reliability. Such an optimization process, however, is constrained by many factors, including, among others: public safety, worker safety, steam cycle efficiency, reliability, and cost. As these factors impose conflicting requirements, there is a need to find an optimum design solution, i.e., one that satisfies all requirements, but not necessarily each requirement optimally. The SEAFP reactor study developed helium-cooled and water-cooled models for assessment purposes. Among other things, the current study demonstrates that neither model offers an optimum solution. Helium cooling offers a high steam cycle efficiency but poor reliability for the cooling of high heat flux components (divertor and first wall). Alternatively, water cooling offers a low steam cycle efficiency, but reasonable reliability for the cooling of such components. It is concluded that an optimum solution includes helium cooling of low heat flux components and water cooling of high heat flux components. Relative to the SEAFP helium model, this hybrid system enhances safety and reliability, while retaining the high steam cycle efficiency of that model.

Power density effect on feasibility of water cooled thorium breeder reactor

International Nuclear Information System (INIS)

Sidik, Permana; Takaki, Naoyuki; Sekimoto, Hiroshi

2008-01-01

Breeding is made possible by the high value of neutron regeneration ratio η for 233 U in thermal energy region. The reactor is fueled by 233 U-Th oxide and it has used the light water as moderator. Some characteristics such as spectrum, η value, criticality, breeding performance and number density are evaluated. Several power densities are evaluated in order to analyze its effect to the breeding performance. The η value of fissile 233 U obtains higher value than 2 which may satisfy the breeding capability especially for thermal reactor for all investigated MFR. The increasing enrichment and decreasing conversion ratio are more significant for MFR 233 U enrichment. Number density of 233 Pa decreases significantly with decreasing power density which leads the reactor has better breeding performance because lower capture rate of 233 Pa. (author)

Gravity Scaling of a Power Reactor Water Shield

International Nuclear Information System (INIS)

Reid, Robert S.; Pearson, J. Boise

2008-01-01

Water based reactor shielding is being considered as an affordable option for potential use on initial lunar surface reactor power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxillary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2006). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa n . These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined

MIF-SCD computer code for thermal hydraulic calculation of supercritical water cooled reactor core

International Nuclear Information System (INIS)

Galina P Bogoslovskaia; Alexander A Karpenko; Pavel L Kirillov; Alexander P Sorokin

2005-01-01

Full text of publication follows: Supercritical pressure power plants constitute the basis of heat power engineering in many countries to day. Starting from a long-standing experience of their operation, it is proposed to develop a new type of fast breeder reactor cooled by supercritical water, which enables the economical indices of NPP to be substantially improved. In the Thermophysical Department of SSC RF-IPPE, an attempt is made to provide thermal-hydraulic validation of the reactor under discussion. The paper presents the results of analysis of the thermal-hydraulic characteristics of fuel subassemblies cooled by supercritical water based on subchannel analysis. Modification of subchannel code MIF - MIF-SCD Code - developed in the SSC RF IPPE is designed as block code and permits one to calculate the coolant temperature and velocity distributions in fuel subassembly channels, the temperature of fuel pin claddings and fuel subassembly wrapper under conditions of irregular geometry and non-uniform axial and radial power generation. The thermal hydraulics under supercritical pressure of water exhibits such peculiarities as abrupt variation of the thermal physical properties in the range of pseudo-critical temperature, the absence of such phenomenon as the critical heat flux which can lead to fuel element burnout in WWERs. As compared with subchannel code for light water, in order to take account of the variation of the coolant properties versus temperature in more detail, a block for evaluating the thermal physical properties of supercritical water versus the local coolant temperature in the fuel subassembly channels was added. The peculiarities of the geometry and power generation in the fuel subassembly of the supercritical reactor are considered as well in special blocks. The results of calculations have shown that considerable preheating of supercritical coolant (several hundreds degrees) can occur in the fuel subassembly. The test calculations according to

Method of cooling a pressure tube type reactor

International Nuclear Information System (INIS)

Kanazawa, Nobuhiro.

1983-01-01

Purpose: To improve the operation efficiency of a nuclear reactor by carrying out cooling depending on the power distribution in the reactor core. Constitution: Reactor core channels are divided into a plurality of channel groups depending on the reactor power, and a water drum and a pump are disposed to each of the channel groups so as to increase the amount of coolants in response to the magnitude of the power from each of the channel groups. In this way, the minimum limiting power ratio can be increased. (Seki, T.)

Thermohydraulic relationships for advanced water cooled reactors and the role of the IAEA

International Nuclear Information System (INIS)

Badulescu, A.; Groeneveld, D.C.

2000-01-01

Under the auspices of the International Atomic Energy Agency (IAEA) a Coordinated Research Program (CRP) on Thermohydraulic Relationships for Advanced Water-Cooled Reactors was carried out from 1995-1998. It was included into the IAEA's Programme following endorsement in 1995 by the International Working Group on Advanced Technologies for Water Cooled Reactors. The overall goal was to promote International Information exchange and cooperation in establishing a consistent set of thermohydraulic relationships that are appropriate for use in analyzing the performance and safety of advanced water-cooled reactors. (authors)

Onsite nondestructive examination techniques for irradiated water-cooled power reactor fuel

International Nuclear Information System (INIS)

1981-03-01

The International Atomic Energy Agency, in response to the recommendations from several Member States, has prepared this Guidebook on Onsite Non-Destructive Techniques for Irradiated Water-Cooled Power Reactor Fuel with the assistance of a number of experts and organizations in this field. During the preparation of this report it became evident that a comparison between different techniques is a most difficult task and depends on a number of factors related to fuel design, plant characteristics and operating conditions. Consequently the emphasis of the report is on the survey of different techniques presently available. It is also to be noted that because the degree of development for any given technique varies significantly among organizations, it is understood that the report should not be used as consensus standard of the minimum capabilities for each class of techniques, nor does it give recommendations in the regulatory sense. Furthermore, the inclusion of some commercial pieces of equipment, services and other products are for illustrative purposes only and neither implies any preference by the Agency nor can the Agency be liable for any material presented in the report

Fast reactor cooled by supercritical light water

Energy Technology Data Exchange (ETDEWEB)

Ishiwatari, Yuki; Mukouhara, Tami; Koshizuka, Seiichi; Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Lab., Tokai, Ibaraki (Japan)

2001-09-01

This report introduces the result of a feasibility study of a fast reactor cooled by supercritical light water (SCFR) with once-through cooling system. It is characterized by (1) no need of steam separator, recirculation system, or steam generator, (2) 1/7 of core flow rate compared with BWR or PWR, (3) high temperature and high pressure permits small turbine and high efficiency exceeding 44%, (4) structure and operation of major components are already experienced by LWRs or thermal power plants. Modification such as reducing blanket fuels and increasing seed fuels are made to achieve highly economic utilization of Pu and high power (2 GWe). The following restrictions were satisfied. (1) Maximum linear heat rate 39 kW/m, (2) Maximum surface temperature of Inconel cladding 620degC, (3) Negative void reactivity coefficient, (4) Fast neutron irradiation rate at the inner surface of pressure vessel less than 2.0x10{sup 19} n/cm{sup 2}. Thus the high power density of 167 MW/m{sup 3} including blanket is thought to contributes economy. The high conversion is attained to be 0.99 Pu fission residual rate by the outer radius of fuel rod of 0.88 mm. The breeding of 1.034 by Pu fission residual rate can be achieved by using briquette (tube-in-shell) type fuel structure. (K. Tsuchihashi)

Gas-Cooled Reactors: the importance of their development

International Nuclear Information System (INIS)

Kasten, P.R.

1978-01-01

Gas-Cooled Reactors are considered to have a significant future impact on the application of fission energy. The specific types are the steam-cycle High-Temperature Gas-Cooled Reactor, the Gas-Cooled Fast Breeder Reactor, the gas-turbine HTGR, and the Very High-Temperature Process Heat Reactor. The importance of developing the above systems is discussed relative to alternative fission power systems involving Light Water Reactors, Heavy Water Reactors, Spectral Shift Controlled Reactors, and Liquid-Metal-Cooled Fast Breeder Reactors. A primary advantage of developing GCRs as a class lies in the technology and cost interrelations, permitting cost-effective development of systems having diverse applications. Further, HTGR-type systems have highly proliferation-resistant characteristics and very attractive safety features. Finally, such systems and GCFRs are mutally complementary. Overall, GCRs provide interrelated systems that serve different purposes and needs; their development can proceed in stages that provide early benefits while contributing to future needs. It is concluded that the long-term importance of the various GCRs is as follows: HTGR, providing a technology for economic GCFRs and HTGR-GTs, while providing a proliferation-resistant reactor system having early economic and fuel utilization benefits; GCFR, providing relatively low cost fissile fuel and reducing overall separative work needs at capital costs lower than those for LMFBRs; HTGR-GT (in combination with a bottoming cycle), providing a very high thermal efficiency system having low capital costs and improved fuel utilization and technology pertinent to VHTRs; HTGR-GT, providing a power system well suited for dry cooling conditions for low-temperature process heat needs; and VHTR, providing a high-temperature heat source for hydrogen production processes

Water chemical control of the TRIGA IPR-R1 reactor primary cooling system

International Nuclear Information System (INIS)

Auler, Lucia M.L.A; Chaves, Renata D.A.; Palmieri, Helena E.L.; Menezes, Maria Angela de B.C.; Oliveira, Paulo F.; Kastner, Geraldo F.; Damazio, Ilza; Fagundes, Oliene dos R.; Cintra, Maria Olivia C.; Andrade, Geraldo V. de; Amaral, Angela M.; Franco, Milton B.; Fortes, Flavio; Gomes, Nilton Carlos; Vidal, Andrea; Maretti Junior, Fausto; Knupp, Eliana A.N.; Souza, Wagner de; Guedes, Joao B.; Furtado, Renato C.S.

2013-01-01

The TRIGA Mark I IPR-R1 reactor located at CDTN/CNEN has been in operation and contributed to research and with services to society since 1960. Is has been used in several activities such as nuclear power plant operation, graduate and post-graduate training courses, isotope production, and as an analytical irradiation tool of different types of samples. Among the several structural and operational safety requirements is the chemical quality control of the primary circuit cooling water. The aim of this work was to check the cooling water quality from the pool reactor. A water sampling plan was proposed (May, 2011 - June, 2012) and presents the results obtained in this period. The natural radioactivity level as gross alpha and gross beta activity and other chemical parameters (pH and electric conductivity) of the samples were analyzed. Some instrumental techniques were used: potentiometric methods (pH), conductometric methods (electrical conductivity, EC) and gross α and gross β proportional counting system). (author)

SBWR: A simplified boiling water reactor

International Nuclear Information System (INIS)

Duncan, J.D.; Sawyer, C.D.; Lagache, M.P.

1987-01-01

An advanced light water reactor concept is being developed for possible application in the 1990's. The concept, known as SBWR is a boiling water reactor which uses natural circulation to provide flow to the reactor core. In an emergency, a gravity driven core cooling system is used. The reactor is depressurized and water from an elevated suppression pool flows by gravity to the reactor vessel to keep the reactor core covered. The concept also features a passive containment cooling system in which water flows by gravity to cool the suppression pool wall. No operator action is required for a period of at least three days. Use of these and other passive systems allows the elimination of emergency diesel generators, core cooling pumps and heat removal pumps which is expected to simplify the plant design, reduce costs and simplify licensing. The concept is being developed by General Electric, Bechtel and the Massachusetts Institute of Technology supported by the Electric Power Research Institute and the United States Department of Energy in the United States. In Japan, The Japan Atomic Power Company has a great interest in this concept

A passive emergency heat sink for water-cooled reactors with particular application to CANDU reactors

International Nuclear Information System (INIS)

Spinks, N.J.

1996-01-01

Water in an overhead pool can serve as a general-purpose passive emergency heat sink for water-cooled reactors. It can be used for containment cooling, for emergency depressurization of the heat transport-system, or to receive any other emergency heat, such as that from the CANDU moderator. The passive emergency water system provides in-containment depressurization of steam generators and no other provision is needed for supply of low-pressure emergency water to the steam generators. For containment cooling, the pool supplies water to the tube side of elevated tube banks inside containment. The elevation with respect to the reactor heat source maximizes heat transport, by natural convection, of hot containment gases. This effective heat transport combines with the large heat-transfer coefficients of tube banks, to reduce containment overpressure during accidents. Cooled air from the tube banks is directed past the break in the heat-transport system, to facilitate removal of hydrogen using passive catalytic recombiners. (author)

Topics to be covered in safety analysis reports for nuclear power plants with pressurized water reactors or boiling water reactors in the F.R.G

International Nuclear Information System (INIS)

Kohler, H.A.G.

1977-01-01

This manual aims at defining the standards to be used in Safety Analysis Reports for Nuclear Power Plants with Pressurized Water Reactors or Boiling Water Reactors in the Federal Republic of Germany. The topics to be covered are: Information about the site (geographic situation, settlement, industrial and military facilities, transport and communications, meteorological conditions, geological, hydrological and seismic conditions, radiological background), description of the power plant (building structures, safety vessel, reactor core, cooling system, ventilation systems, steam power plant, electrical facilities, systems for measurement and control), indication of operation (commissioning, operation, safety measures, radiation monitoring, organization), incident analysis (reactivity incidents, loss-of-coolant incidents, external impacts). (HP) [de

Severe water ingress accident analysis for a Modular High Temperature Gas Cooled Reactor

International Nuclear Information System (INIS)

Zhang Zuoyi; Scherer, Winfried

1997-01-01

This paper analyzes the severe water ingress accidents in the SIEMENS 200MW Modular High Temperature Gas Cooled Reactor (HTR-Module) under the assumption of no active safety protection systems in order to find the safety margin of the current HTR-Module design. A water, steam and helium multi-phase cavity model is originally developed and implemented in the DSNP simulation system. The developed DSNP system is used to simulate the primary circuit of HTR-Module power plant. The comparisons of the models with the TINTE calculations validate the current simulation. After analyzing the effects of blower separation on water droplets, the wall heat storage, etc., it is found that the maximum H 2 O density increase rate in the reactor core is smaller than 0.3 kg/(m 3 s). The liquid water vaporization in the steam generator and H 2 O transport from the steam generator to the reactor core reduces the impulse of the H 2 O in the reactor core. The nuclear reactivity increase caused by the water ingress leads to a fast power excursion, which, however, is inherently counterbalanced by negative feedback effects. Concerning the integrity of the fuel elements, the safety relevant temperature limit of 1600degC was not reached in any case. (author)

Design measures to facilitate implementation of safeguards at future water cooled nuclear power plants

International Nuclear Information System (INIS)

1999-01-01

The report is intended to present guidelines to the State authorities, designers and prospective purchasers of future water cooled power reactors which, if taken into account, will minimize the impact of IAEA safeguards on plant operation and ensure efficient and effective acquisition of safeguards data to the mutual benefit of the Member State, the plant operator and the IAEA. These guidelines incorporate the IAEA's experience in establishing and carrying out safeguards at currently operating nuclear power plants, the ongoing development of safeguards techniques and feedback of experience from plant operators and designers on the impact of IAEA safeguards on plant operation. The following main subjects are included: The IAEA's safeguards function for current and future nuclear power plants; summary of the political and legal foundations of the IAEA's safeguards system; the technical objective of safeguards and the supply and use of required design information; safeguards approaches for nuclear power plants; design implications of experience in safeguarding nuclear power plants and guidelines for future water cooled reactors to facilitate the implementation of safeguards

High temperature on-line monitoring of water chemistry and corrosion control in water cooled power reactors. Report of a co-ordinated research project 1995-1999

International Nuclear Information System (INIS)

2002-07-01

This report documents the results of the Co-ordinated Research Project (CRP) on High Temperature On-line Monitoring of Water Chemistry and Corrosion in Water Cooled Power Reactors (1995-1999). This report attempts to provide both an overview of the state of the art with regard to on-line monitoring of water chemistry and corrosion in operating reactors, and technical details of the important contributions made by programme participants to the development and qualification of new monitoring techniques. The WACOL CRP is a follow-up to the WACOLIN (Investigations on Water Chemistry Control and Coolant Interaction with Fuel and Primary Circuit Materials in Water Cooled Power Reactors) CRP conducted by the IAEA from 1986 to 1991. The WACOLIN CRP, which described chemistry, corrosion and activity-transport aspects, clearly showed the influence of water chemistry on corrosion of both fuel and reactor primary-circuit components, as well as on radiation fields. It was concluded that there was a fundamental need to monitor water-chemistry parameters in real time, reliably and accurately. The objectives of the WACOL CRP were to establish recommendations for the development, qualification and plant implementation of methods and equipment for on-line monitoring of water chemistry and corrosion. Chief investigators from 18 organizations representing 15 countries provided a variety of contributions aimed at introducing proven monitoring techniques into plants on a regular basis and filling the gaps between plant operator needs and available monitoring techniques. The CRP firmly demonstrated that in situ monitoring is able to provide additional and valuable information to plant operators, e.g. ECP, high temperature pH and conductivity. Such data can be obtained promptly, i.e. in real time and with a high degree of accuracy. Reliable techniques and sensor devices are available which enable plant operators to obtain additional information on the response of structural materials in

Method of avoiding hazards resulting from accidents in water-cooled nuclear reactors

International Nuclear Information System (INIS)

Dorner, S.; Schretzmann, K.; Schumacher, G.

1984-01-01

In water-cooled reactors, e.g. BWRs and PWRs, elemental hydrogen is released by hydrolysis (in-leakage). In case of an accident in these reactors or at emergency cooling of e.g., a gas-cooled reactor with water additional hydrogen is produced by chemical reactions of the water with the cladding material. In order to prevent hydrogen pressurizing and the formation of a detonating gas mixture, dry powder containers are provided for in the endangered compartments of the reactor. In case of danger powdered CuO, MnO 2 , Fe 2 O 3 , or CdO, the oxygen content of which recombines with the hydrogen, is ejected from them. In addition, an extinguishing substance with an anticatalytic resp. inhibition effect and/or an inert gas of the group N 2 , He, Ar, CO 2 may be admixed to the powder resp. powder mixture. (orig./PW)

Feasibility of maintaining natural convection mode core cooling in research reactor power upgrades

International Nuclear Information System (INIS)

Ha, J.J.; Belhadj, M.; Aldemir, T.; Christensen, R.N.

1987-01-01

Two operational concerns for natural convection coooled research reactors using plate type fuels are: 1) pool top 16 N activity (PTNA), and 2) nucleate boiling in core channels. The feasibility assessment of a power upgrade while maintaining natural convection mode core cooling requires addressing these operational concerns. Previous studies have shown that: a) The conventional technique for reducing PTNA by plume dispersion may not be effective in a large power upgrade of research reactors with small pools. b) Currently used correlations to predict onset of nucleate boiling (ONB) in thin, rectangular core channels are not valid for low-velocity, upward flows such as encountered in natural convection cooling. The PTNA depends on the velocity distribution in the reactor pool. COMMIX-1A code is used to determine the three-dimensional velocity fields in The Ohio State University Research Reactor (OSURR) pool as a function of varying design conditions, following a power upgrade to 500 kW with LEU fuel. It is shown that a sufficiently deep stagnant water layer can be created below the pool top by properly choosing the disperser flow rate. The ONB heat flux is experimentally determined for channel gaps and upward flow velocities in the range 2mm-4mm and 3-16 cm/sec., respectively. Two alternatives to plume dispersion for reducing PTNA and a new correlation to determine the ONB heat flux in thin, rectangular channels under low-velocity, upward flow conditions are proposed. (Author)

Cooling of pressurized water nuclear reactor vessels

International Nuclear Information System (INIS)

Curet, H.D.

1978-01-01

The improvement of pressurized water nuclear reactor vessels comprising flow dividers providing separate and distinct passages for the flow of core coolant water from each coolant water inlet, the flow dividers being vertically disposed in the annular flow areas provided by the walls of the vessel, the thermal shield (if present), and the core barrel is described. In the event of rupture of one of the coolant water inlet lines, water, especially emergency core coolant water, in the intact lines is thus prevented from by-passing the core by circumferential flow around the outermost surface of the core barrel and is instead directed so as to flow vertically downward through the annulus area between the vessel wall and the core barrel in a more normal manner to increase the probability of cooling of the core by the available cooling water in the lower plenum, thus preventing or delaying thermal damage to the core, and providing time for other appropriate remedial or damage preventing action by the operator

Power oscillation and stability in water cooled reactors

International Nuclear Information System (INIS)

Por, G.; Kis, G.

1998-01-01

Periodic oscillation in measured temperature fluctuation was observed near to surface of a heated rod in certain heat transfer range. The frequency of the peak found in power spectral density of temperature fluctuation and period estimated from the cross correlation function for two axially placed thermocouples change linearly with linear energy (or surface heat) production. It was concluded that a resonance of such surface (inlet) temperature oscillation with the pole of the reactor transfer function can be responsible for power oscillation in BWR and PWR, thus instability is not solely due to reactor transfer function. (author)

Neutronic and thermal hydraulic assessment of fast reactor cooling by water of super critical parameters

International Nuclear Information System (INIS)

Baranaev, Yu. D.; Glebov, A. P.; Ukraintsev, V. F.; Kolesov, V. V.

2007-01-01

Necessity of essential improvement of competitiveness for reactors on light water determines development of new generation power reactors on water of super critical parameters. The main objective of these projects is reaching of high efficiency coefficients while decreasing of investment to NPP and simplification of thermal scheme and high safety level. International programme of IV generation in which super critical reactors present is already started. In the frame of this concept specific Super Critical Fast Reactor with tight lattice of pitch is developing by collaboration of the FEI and IATE. In present article neutronic and thermal hydraulic assessment of fast reactor with plutonium MOX fuel and a core with a double-path of super critical water cooling is presented (SCFR-2X). The scheme of double path of coolant via the core in which the core is divided by radius on central and periphery parts with approximately equal number of fuel assemblies is suggested. Periferia part is cooling while down coming coolant movement. At the down part of core into the mix chamber flows from the periphery assemblies joining and come to the inlet of the central part which is cooling by upcoming flow. Eight zone of different content of MOX fuel are used (4 in down coming and 4 in upcoming) sub zones. Calculation of fuel burn-up and approximate scheme of refueling is evaluated. Calculation results are presented and discussed

Fundamental design bases for independent core cooling in Swedish nuclear power reactors

International Nuclear Information System (INIS)

Jelinek, Tomas

2015-01-01

New regulations on design and construction of nuclear power plants came into force in 2005. The need of an independent core cooling system and if the regulations should include such a requirement was discussed. The Swedish Radiation Safety authority (SSM) decided to not include such a requirement because of open questions about the water balance and started to investigate the consequences of an independent core cooling system. The investigation is now finished and SSM is also looking at the lessons learned from the accident in Fukushima 2011. One of the most important measures in the Swedish national action plan is the implementation of an independent core cooling function for all Swedish power plants. SSM has investigated the basic design criteria for such a function where some important questions are the level of defence in depth and the acceptance criteria. There is also a question about independence between the levels of defence in depth that SSM have included in the criteria. Another issue that has to be taken into account is the complexity of the system and the need of automation where independence and simplicity are very strong criteria. In the beginning of 2014 a memorandum was finalized regarding fundamental design bases for independent core cooling in Swedish nuclear power reactors. A decision based on this memorandum with an implementation plan will be made in the first half of 2014. Sweden is also investigating the possibility to have armed personnel on site, which is not allowed currently. The result from the investigation will have impact on the possibility to use mobile equipment and the level of protection of permanent equipment. In this paper, SSM will present the memorandum for design bases for independent core cooling in Swedish nuclear power reactors that was finalized in March 20147 that also describe SSM's position regarding independence and automation of the independent core cooling function. This memorandum describes the Swedish

Water cooled type nuclear power plant

International Nuclear Information System (INIS)

Arai, Shigeki.

1981-01-01

Purpose: To construct high efficiency a PWR type nuclear power plant with a simple structure by preparing high temperature and pressure water by a PWR type nuclear reactor and a pressurizer, converting the high temperature and high pressure water into steam with a pressure reducing valve and introducing the steam into a turbine, thereby generating electricity. Constitution: A pressurizer is connected downstream of a PWR type nuclear reactor, thereby maintaining the reactor at high pressure. A pressure-reducing valve is provided downstream of the pressurizer, the high temperature and pressure water is reduced in pressure, thereby producing steam. The steam is fed to a turbine, and electric power is generated by a generator connected to the turbine. The steam exhausted from the turbine is condensed by a condenser into water, and the water is returned through a feedwater heater to the reactor. Since the high temperature and pressure water in thus reduced in pressure thereby evaporating it, the steam can be more efficiently produced than by a steam generator. (Sekiya, K.)

Recent IAEA activities to support advanced water cooled reactor technology development

International Nuclear Information System (INIS)

Choi, J.-H.; Bilbao y Leon, S.; Rao, A.S.

2009-01-01

The International Atomic Energy Agency (IAEA) is the world's center of cooperation in the nuclear field. The IAEA works with its Member States and multiple partners worldwide to promote safe, secure and peaceful nuclear technologies. To catalyse innovation in nuclear power technology in Member States, the IAEA coordinates cooperative research, promotes information exchange, and analyses technical data and results, with a focus on reducing capital costs and construction periods while further improving performance, safety and proliferation resistance. This paper summarizes the recent major IAEA activities to support technology development for water cooled reactors, which is the most common type of reactor design at present and will probably still be in the near future. (author)

High power density reactors based on direct cooled particle beds

International Nuclear Information System (INIS)

Powell, J.R.; Horn, F.L.

1985-01-01

Reactors based on direct cooled HTGR type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out long the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBR's) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed. 12 figs

Power Conversion Study for High Temperature Gas-Cooled Reactors

International Nuclear Information System (INIS)

Chang Oh; Richard Moore; Robert Barner

2005-01-01

The Idaho National Laboratory (INL) is investigating a Brayton cycle efficiency improvement on a high temperature gas-cooled reactor (HTGR) as part of Generation-IV nuclear engineering research initiative. There are some technical issues to be resolved before the selection of the final design of the high temperature gas cooled reactor, called as a Next Generation Nuclear Plant (NGNP), which is supposed to be built at the INEEL by year 2017. The technical issues are the selection of the working fluid, direct vs. indirect cycle, power cycle type, the optimized design in terms of a number of intercoolers, and others. In this paper, we investigated a number of working fluids for the power conversion loop, direct versus indirect cycle, the effect of intercoolers, and other thermal hydraulics issues. However, in this paper, we present part of the results we have obtained. HYSYS computer code was used along with a computer model developed using Visual Basic computer language

Experience in the development of metal uranium-base nuclear fuel for heavy-water gas-cooled reactors

International Nuclear Information System (INIS)

Ashikhmin, V.P.; Vorob'ev, M.A.; Gusarov, M.S.; Davidenko, A.S.; Zelenskij, V.F.; Ivanov, V.E.; Krasnorutskij, V.S.; Petel'guzov, I.A.; Stukalov, A.I.

1978-01-01

Investigations were carried out to solve the problem of making the development of radiation-resistant uranium fuel for power reactors including the heavy-water gas-cooled KS-150 reactor. Factors are considered that limit the lifetime of uranium fuel elements, and the ways of suppressing them are discussed. Possible reasons of the insufficient radiation resistance of uranium rod fuel element and the progress attained are analyzed. Some general problems on the fuel manufacture processes are discussed. The main results are presented on the operation of the developed fuel in research reactor loops and the commercial heavy-water KS-150 reactor. The results confirm an exceptionally high radiation resistance of fuel to burn-ups of 1.5-2%. The successful solution of a large number of problems associated with the development of metal uranium fuel provides for new possibilities of using metal uranium in power reactors

Progress in development and design aspects of advanced water cooled reactors

International Nuclear Information System (INIS)

1992-12-01

The objective of the Technical Committee Meeting (TCM) was to provide an international forum for technical specialists to review and discuss technology developments and design work for advanced water cooled reactors, safety approaches and features of current water cooled reactors and to identify, understand and describe advanced features for safety and operational improvements. The TCM was attended by 92 participants representing 18 countries and two international organizations and included 40 presentations by authors of 14 countries and one international organization. A separate abstract was prepared for each of these presentations. Refs, figs, tabs

Cooling and heating facility for nuclear power plant

International Nuclear Information System (INIS)

Kakuta, Atsuro

1994-01-01

The present invention concerns a cooling and heating facility for a nuclear power plant. Namely, a cooling water supply system supplies cooling water prepared by a refrigerator for cooling the inside of the plant. A warm water supply system supplies warm water having its temperature elevated by using an exhausted heat from a reactor water cleanup system. The facility comprises a heat pump-type refrigerator disposed in a cold water supply system for producing cold water and warm water, and warm water pipelines for connecting the refrigerator and the warm water supply system. With such a constitution, when the exhaust heat from the reactor water cleanup system can not be used, warm water prepared by the heat pump type refrigerator is supplied to the warm water supply system by way of the warm water pipelines. Accordingly, when the exhaust heat from the reactor water cleanup system can not be used such as upon inspection of the plant, a portion of the refrigerators in a not-operated state can be used for heating. Supply of boiler steams in the plant is no more necessary or extremely reduced. (I.S.)

Aiming at super long term application of nuclear energy. Scope and subjects on the water cooled breeder reactor, the 'reduced moderation water reactor'

International Nuclear Information System (INIS)

Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji

2001-01-01

In order to make possible on nuclear energy application for super long term, development of sodium cooling type fast breeder reactor (FBR) has been carried out before today. However, as it was found that its commercialization was technically and economically difficult beyond expectation, a number of nations withdrew from its development. And, as Japan has continued its development, scope of its actual application is not found yet. Now, a research and development on a water cooling type breeder reactor, the reduced moderation water reactor (RMWR)' using LWR technology has now been progressed under a center of JAERI. This RMWR is a reactor intending a jumping upgrade of conversion ratio by densely arranging fuel bars to shift neutron spectrum to faster region. The RMWR has a potential realizable on full-dress plutonium application at earlier timing through its high conversion ratio, high combustion degree, plutonium multi-recycling, and so on. And, it has also feasibility to solve uranium resource problem by realization of conversion ratio with more than 1.0, to contribute to super long term application of nuclear energy. Here was investigated on an effect of reactor core on RMWR, especially of its conversion ratio and plutonium loading on introduction effect as well as on how RMWR could be contributed to reduction of uranium resource consumption, by drawing some scenario on development of power generation reactor and fuel cycle in Japan under scope of super long term with more than 100 years in future. And, trial calculation on power generation cost of the RMWR was carried out to investigate some subjects at a viewpoint of upgrading on economy. (G.K.)

Safety analysis for K reactor and impact of cooling tower installation

International Nuclear Information System (INIS)

Fields, C.C.; Wooten, L.A.; Geeting, M.W.; Morgan, C.E.; Buczek, J.A.; Smith, D.C.

1993-01-01

This paper describes the safety analysis of the Savannah River site K-reactor loss-of-cooling-water-supply (LOCWS) event and the impact on the analysis of a natural-draft cooling tower, which was installed in 1992. Historically (1954 to 1992), the K-reactor secondary cooling system [called the cooling water system (CWS)] used water from the Savannah River pumped to a 25-million-gal basin adjacent to the reactor. Approximately 170 000 gal/min were pumped from the basin through heat exchangers to remove heat from the primary cooling system. This water then entered a smaller basin, where it flowed over a weir and eventually returned to the Savannah River. The 25-million-gal basin is at a higher elevation than the heat exchangers and the smaller basin to supply cooling by gravity flow (which is sufficient to remove decay heat) if power to the CWS pumps is interrupted. Small amounts of cooling water are also used for other essential equipment such as diesels, motors, and oil coolers. With the cooling tower installed, ∼85% of the cooling water flows from the small basin by gravity to the cooling tower instead of returning to the Savannah River. After being cooled, it is pumped back to the 25-million-gal basin. River water is supplied only to make up for evaporation and the blowdown stream

Passive safety features in current and future water cooled reactors

International Nuclear Information System (INIS)

1990-11-01

Better understanding of the passive safety systems and components in current and future water-cooled reactors may enhance the safety of present reactors, to the extend passive features are backfitted. This better understanding should also improve the safety of future reactors, which can incorporate more of these features. Passive safety systems and components may help to prevent accidents, core damage, or release radionuclides to the environment. The Technical Committee Meeting which was hosted by the USSR State Committee for Utilization of Nuclear Energy was attended by about 80 experts from 16 IAEA Member States and the NEA-OECD. A total of 21 papers were presented during the meeting. The objective of the meeting was to review and discuss passive safety systems and features of current and future water cooled reactor designs and to exchange information in this area of activity. A separate abstract was prepared for each of the 21 papers published in this proceedings. Refs, figs and tabs

Assessments of Water Ingress Accidents in a Modular High-Temperature Gas-Cooled Reactor

International Nuclear Information System (INIS)

Zhang Zuoyi; Dong Yujie; Scherer, Winfried

2005-01-01

Severe water ingress accidents in the 200-MW HTR-module were assessed to determine the safety margins of modular pebble-bed high-temperature gas-cooled reactors (HTR-module). The 200-MW HTR-module was designed by Siemens under the criteria that no active safety protection systems were necessary because of its inherent safe nature. For simulating the behavior of the HTR-module during severe water ingress accidents, a water, steam, and helium multiphase cavity model was developed and implemented in the dynamic simulator for nuclear power plants (DSNP) simulation system. Comparisons of the DSNP simulations incorporating these models with experiments and with calculations using the time-dependent neutronics and temperature dynamics code were made to validate the simulation. The analysis of the primary circuit showed that the maximum water concentration increase in the reactor core was 3 s). The water vaporization in the steam generator and characteristics of water transport from the steam generator to the reactor core would reduce the rate of water ingress into the reactor core. The analysis of a full cavitation of the feedwater pump showed that if the secondary circuit could be depressurized, the feedwater pump would be stopped by the full cavitation. This limits the water transported from the deaerator to the steam generator. A comprehensive simulation of the HTR-module power plant showed that the water inventory in the primary circuit was limited to ∼3000 kg. The nuclear reactivity increase caused by the water ingress would lead to a fast power excursion, which would be inherently counterbalanced by negative feedback effects. The integrity of the fuel elements, because the safety-relevant temperature limit of 1600 deg. C is not reached in any case, is not challenged

Auxiliary water supply device for BWR type reactor

International Nuclear Information System (INIS)

Sasagawa, Hiroshi.

1994-01-01

In the device of the present invention, a cooling condensation means is disposed to a steam discharge channel of a turbine for driving pumps to directly return condensates to the reactor, so that the temperature of the suppression pool water is not elevated. Namely, the cooling condensation means for discharged steams is disposed to the discharge channel of the turbine. The condensate channel from the cooling condensation means is connected to a sucking side of the turbine driving pump. With such a constitution, when the reactor is isolated from a main steam system, reactor scram is conducted. Although the reactor water level is lowered by the reactor scram, the lowering of the reactor water level is prevented by supplementing cooling water by the turbine driving pump using steams generated in the reactor as a power source. The discharged steams after driving the turbine are cooled and condensated by the cooling condensation means by way of the discharge channel and returned to the reactor again by way of the condensate channel. With such procedures, since the temperature of suppression pool water is not elevated, there is no need to operate other cooling systems. In addition, auxiliary water can be supplied for a long period of time. (I.S.)

Safety analysis of a high temperature supercritical pressure light water cooled and moderated reactor

International Nuclear Information System (INIS)

Ishiwatari, Y.; Oka, Y.; Koshizuka, S.

2002-01-01

A safety analysis code for a high temperature supercritical pressure light water cooled reactor (SCLWR-H) with water rods cooled by descending flow, SPRAT-DOWN, is developed. The hottest channel, a water rod, down comer, upper and lower plenums, feed pumps, etc. are modeled as junction of nodes. Partial of the feed water flows downward from the upper dome of the reactor pressure vessel to the water rods. The accidents analyzed here are total loss of feed water flow, feed water pump seizure, and control rods ejection. All the accidents satisfy the criteria. The accident event at which the maximum cladding temperature is the highest is total loss of feedwater flow. The transients analyzed here are loss of feed water heating, inadvertent start-up of an auxiliary water supply system, partial loss of feed water flow, loss of offsite power, loss of load, and abnormal withdrawal of control rods. All the transients satisfied the criteria. The transient event for which the maximum cladding temperature is the highest is control rod withdrawal at normal operation. The behavior of loss of load transient is different from that of BWR. The power does not increase because loss of flow occurs and the density change is small. The sensitivities of the system behavior to various parameters during transients and accidents are analyzed. The parameters having strong influence are the capacity of the auxiliary water supply system, the coast down time of the main feed water pumps, and the time delay of the main feed water pumps trip. The control rod reactivity also has strong influence. (authors)

78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

Science.gov (United States)

2013-10-24

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling....79.1, ``Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors.'' This... emergency core cooling systems (ECCSs) for boiling- water reactors (BWRs) whose licenses are issued after...

Improving Safety, Economic, Substantiality, and Security of Nuclear Energy with Canadian Super-Critical Water-cooled Reactor Concept

International Nuclear Information System (INIS)

Hamilton, Holly; Pencer, Jeremy; Yetisir, Metin; Leung, Laurence

2012-01-01

Super-Critical Water-cooled Reactor is one of the six design concepts being developed under the Generation IV International Forum. It is the only concept evolving from the water-cooled reactors and taking advantages of the balance-of-plant design and operation experience of the fossil-power plants. Canada is developing the SCR concept from the well-established pressure-tube reactor technology. The Canadian SCWR maintains modular design approach using relative small fuel channels with the separation of coolant and moderator. It is equipped with an advanced fuel channel design that is capable to transfer decay heat from the fuel to the moderator under the long-term cooling stage. Coupled with the advanced passive-moderator cooling system, cooling of fuel and fuel channel is continuous even without external power or operator intervention. The Canadian SCWR is operating at a pressure of 25 MPa with a core outlet temperature of 625 deg. C. This has led to a drastic increase in thermal efficiency to 48% from 34% of the current fleet of reactors (a 40% rise in relative efficiency). With the high core outlet temperature, a direct thermal cycle has been adopted and has led to simplification in plant design attributing to the cost reduction compared to the current reactor designs. The Canadian SCWR adopts the advanced Thorium fuel cycle to enhance the substantiality, economic, and security. than uranium in the world (estimated to be three times more). This provides the long-term fuel supply. Thorium's price is stable compared to uranium and is consistently lower than uranium. This would maintain the predictability and economic of fuel supply. Thorium itself is a non-fissile material and once irradiated requires special handling. This improves proliferative resistance. The objective of this paper is to highlight these improvements in generating nuclear energy with the Canadian SCWR

Cooling system upon reactor isolation

International Nuclear Information System (INIS)

Yamamoto, Kohei; Oda, Shingo; Miura, Satoshi

1992-01-01

A water level indicator for detecting the upper limit value for a range of using a suppression pool and a thermometer for detecting the temperature of water at the cooling water inlet of an auxiliary device are disposed. When a detection signal is intaken and the water level in the suppression pool reach the upper limit value for the range of use, a secondary flow rate control value is opened and a primary flow rate control valve is closed. When the temperature of the water at the cooling water inlet of the auxiliary device reaches the upper limit value, the primary and the secondary flow rate control valves are opened. During a stand-by state, the first flow rate control valve is set open and the secondary flow rate control valve is set closed respectively. After reactor isolation, if a reactor water low level signal is received, an RCIC pump is actuated and cooling water is sent automatically under pressure from a condensate storage tank to the reactor and the auxiliary device requiring coolants by way of the primary flow rate control valve. Rated flow rate is ensured in the reactor and cooling water of an appropriate temperature can be supplied to the auxiliary device. (N.H.)

Cooling system for auxiliary reactor component

International Nuclear Information System (INIS)

Fujihira, Tomoko.

1991-01-01

A cooling system for auxiliary reactor components comprises three systems, that is, two systems of reactor component cooling water systems (RCCW systems) and a high pressure component cooling water system (HPCCW system). Connecting pipelines having partition valves are intervened each in a cooling water supply pipeline to an emmergency component of each of the RCCW systems, a cooling water return pipeline from the emmergency component of each of the RCCW systems, a cooling water supply pipeline to each of the emmergency components of one of the RCCW system and the HPCCW system and a cooling water return pipeline from each of the emmergency components of one of the RCCW system and the HPCCW system. With such constitution, cooling water can be supplied also to the emmergency components in the stand-by system upon periodical inspection or ISI, thereby enabling to improve the backup performance of the emmergency cooling system. (I.N.)

Environmental effects of large discharges of cooling water. Experiences from Swedish nuclear power plants

International Nuclear Information System (INIS)

Ehlin, Ulf; Lindahl, Sture; Neuman, Erik; Sandstroem, Olof; Svensson, Jonny

2009-07-01

Monitoring the environmental effects of cooling water intake and discharge from Swedish nuclear power stations started at the beginning of the 1960s and continues to this day. In parallel with long-term monitoring, research has provided new knowledge and methods to optimise possible discharge locations and design, and given the ability to forecast their environmental effects. Investigations into the environmental effects of cooling-water are a prerequisite for the issuing of power station operating permits by the environmental authorities. Research projects have been carried out by scientists at universities, while the Swedish Environmental Protection Agency, the Swedish Board of Fisheries, and the Swedish Meteorological and Hydrological Institute, SMHI, are responsible for the greater part of the investigations as well as of the research work. The four nuclear power plants dealt with in this report are Oskarshamn, Ringhals, Barsebaeck and Forsmark. They were taken into operation in 1972, 1975, 1975 and 1980 resp. - a total of 12 reactors. After the closure of the Barsebaeck plants in 2005, ten reactors remain in service. The maximum cooling water discharge from the respective stations was 115, 165, 50 and 135 m 3 /s, which is comparable to the mean flow of an average Swedish river - c:a 150 m 3 /s. The report summarizes studies into the consequences of cooling water intake and discharge. Radiological investigations made at the plants are not covered by this review. The strategy for the investigations was elaborated already at the beginning of the 1960s. The investigations were divided into pre-studies, baseline investigations and monitoring of effects. Pre-studies were partly to gather information for the technical planning and design of cooling water intake and outlet constructions, and partly to survey the hydrographic and ecological situation in the area. Baseline investigations were to carefully map the hydrography and ecology in the area and their natural

Current status of nuclear power generation in Japan and directions in water cooled reactor technology development

International Nuclear Information System (INIS)

Miwa, T.

1991-01-01

Electric power demand aspects and current status of nuclear power generation in Japan are outlined. Although the future plan for nuclear power generation has not been determined yet the Japanese nuclear research centers and institutes are investigating and developing some projects on the next generation of light water reactors and other types of reactors. The paper describes these main activities

Minimization of radioactive material deposition in water-cooled nuclear reactors

International Nuclear Information System (INIS)

Ruiz, C.P.; Blaies, D.M.

1988-01-01

This patent describes the method for inhibiting the deposition of radioactive cobalt in a water-bearing vessel of a water-cooled nuclear reactor which comprises adding zinc ion to water entering the water-bearing vessel. The improvement contains a substantially lower proportion of the /sup 64/Zn isotope than naturally occurring zinc

Auxiliary cooling device for power plant

International Nuclear Information System (INIS)

Yamanoi, Kozo.

1996-01-01

An auxiliary cooling sea water pipeline for pumping up cooling sea water, an auxiliary cooling sea water pipeline and a primary side of an auxiliary cooling heat exchanger are connected between a sea water taking vessel and a sea water discharge pit. An auxiliary cooling water pump is connected to an auxiliary water cooling pipeline on the second side of the auxiliary cooling heat exchanger. The auxiliary cooling water pipeline is connected with each of auxiliary equipments of a reactor system and each of auxiliary equipments of the turbine system connected to a turbine auxiliary cooling water pipeline in parallel. During ordinary operation of the reactor, heat exchange for each of the auxiliary equipments of the reactor and heat exchange for each of the equipments of the turbine system are conducted simultaneously. Since most portions of the cooling devices of each of the auxiliary equipments of the reactor system and each of the auxiliary equipments of the turbine system can be used in common, the operation efficiency of the cooling device is improved. In addition, the space for the pipelines and the cost for the equipments can be reduced. (I.N.)

Emergency Cooling of Nuclear Power Plant Reactors With Heat Removal By a Forced-Draft Cooling Tower

Energy Technology Data Exchange (ETDEWEB)

Murav’ev, V. P., E-mail: murval1@mail.ru

2016-07-15

The feasibility of heat removal during emergency cooling of a reactor by a forced-draft cooling tower with accumulation of the peak heat release in a volume of precooled water is evaluated. The advantages of a cooling tower over a spray cooling pond are demonstrated: it requires less space, consumes less material, employs shorter lines in the heat removal system, and provides considerably better protection of the environment from wetting by entrained moisture.

Thermal-hydraulic simulation and analysis of Research Reactor Cooling Systems

International Nuclear Information System (INIS)

EL Khatib, H.H.A.

2013-01-01

The objective of the present study is to formulate a model to simulate the thermal hydraulic behavior of integrated cooling system in a typical material testing reactor (MTR) under loss of ultimate heat sink, the model involves three interactively coupled sub-models for reactor core, heat exchanger and cooling tower. The developed model predicts the temperature profiles in addition it predicts inlet and outlet temperatures of the hot and cold stream as well as the heat exchangers and cooling tower. The model is validated against PARET code for steady-state operation and also verified by the reactor operational records, and then the model is used to simulate the thermal-hydraulic behavior of the reactor under a loss of ultimate heat sink. The simulation is performed for two operational regimes named regime I of (11 MW) thermal power and three operated cooling tower cells and regime II of (22 MW) thermal power and six operated cooling tower cells. In regime I, the simulation is performed for 1, 2 and 3 cooling tower failed cells while in regime II, it is performed for 1, 2, 3, 4, 5 and 6 cooling tower failed cells. The safety action is conducted by the reactor protection system (RPS) named power reduction safety action, it is triggered to decrease the reactor power by amount of 20% of the present power when the water inlet temperature to the core reaches 43 degree C and a scram (emergency shutdown) is triggered in case of the inlet temperature reaches 44 degree C. The model results are analyzed and discussed. The temperature profiles of fuel, clad and coolant are predicted during transient where its maximum values are far from thermal hydraulic limits.

A passive emergency heat sink for water cooled reactors with particular application to CANDU reg-sign reactors

International Nuclear Information System (INIS)

Spinks, N.J.

1996-01-01

Water in an overhead pool can serve as a general-purpose passive emergency heat sink for water-cooled reactors. It can be used for containment cooling, for emergency depressurization of the heat transport-system, or to receive any other emergency heat, such as that from the CANDU reg-sign moderator. The passive emergency water system provides in-containment depressurization of steam generators and no other provision is needed for supply of low-pressure emergency water to the steam generators. For containment cooling, the pool supplies water to the tube side of elevated tube banks inside containment. The elevation with respect to the reactor heat source maximizes heat transport, by natural convection, of hot containment gases. This effective heat transport combines with the large heat-transfer coefficients of tube banks, to reduce containment overpressure during accidents. Cooled air from the tube banks is directed past the break in the heat-transport system, to facilitate removal of hydrogen using passive catalytic recombiners

WATER BOILER REACTOR

Science.gov (United States)

King, L.D.P.

1960-11-22

As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

Functional systems of a pressurized water reactor

International Nuclear Information System (INIS)

Heinzel, V.

1982-01-01

The main topics, discussed in the present paper, are: - Principle design of the reactor coolant system - reactor pressure vessel with internals - containment design - residual heat removal and emergency cooling systems - nuclear component cooling systems - emergency feed water systems - plant electric power supply system. (orig./RW)

Water and Regolith Shielding for Surface Reactor Missions

Science.gov (United States)

Poston, David I.; Ade, Brian J.; Sadasivan, Pratap; Leichliter, Katrina J.; Dixon, David D.

2006-01-01

This paper investigates potential shielding options for surface power fission reactors. The majority of work is focused on a lunar shield that uses a combination of water in stainless-steel cans and lunar regolith. The major advantage of a water-based shield is that development, testing, and deployment should be relatively inexpensive. This shielding approach is used for three surface reactor concepts: (1) a moderated spectrum, NaK cooled, Hastalloy/UZrH reactor, (2) a fast-spectrum, NaK-cooled, SS/UO2 reactor, and (3) a fast-spectrum, K-heat-pipe-cooled, SS/UO2 reactor. For this study, each of these reactors is coupled to a 25-kWt Stirling power system, designed for 5 year life. The shields are designed to limit the dose both to the Stirling alternators and potential astronauts on the surface. The general configuration used is to bury the reactor, but several other options exist as well. Dose calculations are presented as a function of distance from reactor, depth of buried hole, water boron concentration (if any), and regolith repacked density.

Water and Regolith Shielding for Surface Reactor Missions

International Nuclear Information System (INIS)

Poston, David I.; Sadasivan, Pratap; Dixon, David D.; Ade, Brian J.; Leichliter, Katrina J.

2006-01-01

This paper investigates potential shielding options for surface power fission reactors. The majority of work is focused on a lunar shield that uses a combination of water in stainless-steel cans and lunar regolith. The major advantage of a water-based shield is that development, testing, and deployment should be relatively inexpensive. This shielding approach is used for three surface reactor concepts: (1) a moderated spectrum, NaK cooled, Hastalloy/UZrH reactor, (2) a fast-spectrum, NaK-cooled, SS/UO2 reactor, and (3) a fast-spectrum, K-heat-pipe-cooled, SS/UO2 reactor. For this study, each of these reactors is coupled to a 25-kWt Stirling power system, designed for 5 year life. The shields are designed to limit the dose both to the Stirling alternators and potential astronauts on the surface. The general configuration used is to bury the reactor, but several other options exist as well. Dose calculations are presented as a function of distance from reactor, depth of buried hole, water boron concentration (if any), and regolith repacked density

The low-temperature water-water reactor for district heating atomic power plant (DHPP)

International Nuclear Information System (INIS)

Skvortsov, S.A.; Sokolov, I.N.; Krauze, L.V.; Nikiporetz, Yu.G.; Philimonov, Yu.V.

1977-01-01

The district heating atomic power plant in the article is distinguished by the increased reliability and safety of operation that was provided by the use of following main principles: relatively low parameters of the coolant; the intergral arrangement of equipment and accordingly the minimum branching of the reactor circuit; the natural circulation of coolant of the primary circuit in the steady-state, transient and emergency regimes of reactor operation; the considerable reserves of cold water of the primary circuit in the reactor vessel, providing the emergency cooling; the application of two shells each of which is designed for the total working pressure, the second shell is made of prestressed reinforced concrete that eliminates its brittle failure. (M.S.)

Control of Canadian once-through direct cycle supercritical water-cooled reactors

International Nuclear Information System (INIS)

Sun, Peiwei; Wang, Baosheng; Zhang, Jianmin; Su, Guanghui

2015-01-01

Highlights: • Dynamic characteristics of Canadian SCWR are analyzed. • Hybrid feedforward and feedback control is adopted to deal with cross-coupling. • Gain scheduling control with smooth weight is applied to deal with nonlinearity. • It demonstrates through simulation that the control requirements are satisfied. - Abstract: Canadian supercritical water-cooled reactor (SCWR) can be modelled as a Multiple-input Multiple-output (MIMO) system. It has a high power-to-flow ratio, strong cross-coupling and high degree of nonlinearity in its dynamic characteristics. Among the outputs, the steam temperature is strongly affected by the reactor power and the most challenging to control. It is difficult to adopt a traditional control system design methodology to obtain a control system with satisfactory performance. In this paper, feedforward control is applied to reduce the effect on steam temperature from the reactor power. Single-input Single-output (SISO) feedback controllers are synthesized in the frequency domain. Using the feedforward controller, the steam temperature variation due to disturbances at the reactor power has been significantly suppressed. The control system can effectively maintain the overall system stability and regulate the plant around a specified operating condition. To deal with the nonlinearities, gain scheduling control strategy is adopted. Different sets of controllers combined by smooth weight functions are used for the plant at different load conditions. The proposed control strategies have been evaluated under various operating scenarios. Simulation results show that satisfactory performance can successfully achieved by the designed control system

Chemistry control challenges in a supercritical water-cooled reactor

International Nuclear Information System (INIS)

Guzonas, David; Tremaine, Peter; Jay-Gerin, Jean-Paul

2009-01-01

The long-term viability of a supercritical water-cooled reactor (SCWR) will depend on the ability of designers to predict and control water chemistry to minimize corrosion and the transport of corrosion products and radionuclides. Meeting this goal requires an enhanced understanding of water chemistry as the temperature and pressure are raised beyond the critical point. A key aspect of SCWR water chemistry control will be mitigation of the effects of water radiolysis; preliminary studies suggest markedly different behavior than that predicted from simple extrapolations from conventional water-cooled reactor behavior. The commonly used strategy of adding excess hydrogen at concentrations sufficient to suppress the net radiolytic production of primary oxidizing species may not be effective in an SCWR. The behavior of low concentrations of impurities such as transition metal corrosion products, chemistry control agents, anions introduced via make-up water or from ion-exchange resins, and radionuclides (e.g., 60 Co) needs to be understood. The formation of neutral complexes increases with temperature, and can become important under near-critical and supercritical conditions; the most important region is from 300-450 C, where the properties of water change dramatically, and solvent compressibility effects exert a huge influence on solvation. The potential for increased transport and deposition of corrosion products (active and inactive), leading to (a) increased deposition on fuel cladding surfaces, and (b) increased out-of-core radiation fields and worker dose, must be assessed. There are also significant challenges associated with chemistry sampling and monitoring in an SCWR. The typical methods used in current reactor designs (grab samples, on-line monitors at the end of a cooled, depressurized sample line) will be inadequate, and in-situ measurements of key parameters will be required. This paper describes current Canadian activities in SCWR chemistry and chemistry

Lead- or Lead-bismuth-cooled fast reactors

International Nuclear Information System (INIS)

Bouchter, J.C.; Courouau, J.L.; Dufour, P.; Guidez, J.; Latge, C.; Martinelli, L.; Renault, C.; Rimpault, G.

2014-01-01

Lead-cooled fast reactors are one of the 6 concepts retained for the 4. generation of nuclear reactors. So far no lead-cooled reactors have existed in the world except lead-bismuth-cooled reactors in soviet submarines. Some problems linked to the use of the lead-bismuth eutectic appeared but were satisfactorily solved by a more rigorous monitoring of the chemistry of the lead-bismuth coolant. Lead presents various advantages as a coolant: no reactivity with water and the air,a high boiling temperature and low contamination when irradiated. The main asset of the lead-bismuth alloy is the drop of the fusion temperature from 327 C degrees to 125 C degrees. The main drawback of using lead (or lead-bismuth) is its high corrosiveness with metals like iron, chromium and nickel. The high corrosiveness of the coolant implies low flow velocities which means a bigger core and consequently a bigger reactor containment. Different research programs in the world (in Europe, Russia and the USA) are reviewed in the article but it appears that the development of this type of reactor requires technological breakthroughs concerning materials and the resistance to corrosion. Furthermore the concept of lead-cooled reactors seems to be associated to a range of low output power because of the compromise between the size of the reactor and its resistance to earthquakes. (A.C.)

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

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)

Safety problems of nuclear power plants with channel-type graphite boiling water reactors

International Nuclear Information System (INIS)

Emel'yanov, I.Ya.; Vasilevskij, V.P.; Volkov, V.P.; Gavrilov, P.A.; Kramerov, A.Ya.; Kuznetsov, S.P.; Kunegin, E.P.; Rybakov, N.Z.

1977-01-01

Construction of nuclear power plants in a highly populated region near large industrial centres necessitates to pay a special attention to their nuclear and radiation safety. Safety problems of nuclear reactor operation are discussed, in particular, they are: reliable stoppage of fission chain reaction at any emergency cases; reliable core cooling with failure of various equipment; emergency core cooling with breached pipes of a circulating circuit; and prevention of radioactive coolant release outside the nuclear power plant in amount exceeding the values adopted. Channel-type water boiling reactors incorporate specific features requiring a new approach to safety operation of a reactor and a nuclear power plant. These include primarily a rather large steam volume in the coolant circuit, large amount of accumulated heat, void reactivity coefficient. Channel-type reactors characterized by fair neutron balance and flexible fuel cycle, have a series of advantages alleviating the problem of ensuring their safety. The possibility of reliable control over the state of each channel allows to replace failed fuel elements by the new ones, when operating on-load, to increase the number of circulating loops and reduce the diameter of main pipelines, simplifies significantly the problem of channel emergency cooling and localization of a radioactive coolant release from a breached circuit. The concept of channel-type reactors is based on the solution of three main problems. First, plant safety should be assured in emergency switch off of separate units and, if possible, energy conditions should be maintained, this is of particular importance considering the increase in unit power. Second, the system of safety and emergency cooling should eliminate a great many failures of fuel elements in case of potential breaches of any tube in the circulating circuit. Finally, rugged boxes and localizing devices should be provided to exclude damage of structural elements of the nuclear power

Supercritical Carbon Dioxide turbomachinery design for water-cooled Small Modular Reactor application

International Nuclear Information System (INIS)

Lee, Jekyoung; Lee, Jeong Ik; Yoon, Ho Joon; Cha, Jae Eun

2014-01-01

Highlights: • We described the concept of coupling the S-CO 2 Brayton cycle to the water-cooled SMRs. • We describe a turbomachinery design code called KAISD T MD that can use real gases too. • We suggest changes to the S-CO 2 cycle layout with multiple-independent shafts. • KAIST T MD was used to design the turbomachinery of suggested layout. - Abstract: The Supercritical Carbon Dioxide (S-CO 2 ) Brayton cycle has been gaining attention due to its compactness and high efficiency at moderate turbine inlet temperature. Previous S-CO 2 cycle research works in the field of nuclear engineering were focused on its application to the next generation reactor with higher turbine inlet temperature than the existing conventional water-cooled nuclear power plants. However, it was shown in authors’ previous paper that the advantages of the S-CO 2 Brayton cycle can be also further applied to the water-cooled Small Modular Reactor (SMR) with a success, since SMR requires minimal overall footprint while retaining high performance. One of the major issues in the S-CO 2 Brayton cycle is the selection and design of appropriate turbomachinery for the designed cycle. Because most of the nuclear industry uses incompressible working fluids or ideal gases in the turbomachinery, a more detailed examination of the design of the turbomachinery is required for a power system that uses S-CO 2 as working fluid. This is because the S-CO 2 Brayton cycle high efficiency is the result of the non-ideal variation of properties near the CO 2 critical point. Thus, the major focus of this paper is to suggest the design of the turbomachinery necessary for the S-CO 2 Brayton cycle coupled to water cooled SMRs. For this reason, a S-CO 2 Brayton cycle turbomachinery design methodology was suggested and the suggested design methodology was first tested with the existing experimental data to verify its capability. After then, it was applied to the proposed reference system to demonstrate its

Supercritical Carbon Dioxide turbomachinery design for water-cooled Small Modular Reactor application

Energy Technology Data Exchange (ETDEWEB)

Lee, Jekyoung, E-mail: leejaeky85@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1 Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Lee, Jeong Ik, E-mail: jeongiklee@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1 Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Yoon, Ho Joon, E-mail: hojoon.yoon@kustar.ac.ae [Khalifa University of Science, Technology and Research (KUSTAR), P.O. Box 127788, Abu Dhabi (United Arab Emirates); Cha, Jae Eun, E-mail: jecha@kaeri.re.kr [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

2014-04-01

Highlights: • We described the concept of coupling the S-CO{sub 2} Brayton cycle to the water-cooled SMRs. • We describe a turbomachinery design code called KAISD{sub T}MD that can use real gases too. • We suggest changes to the S-CO{sub 2} cycle layout with multiple-independent shafts. • KAIST{sub T}MD was used to design the turbomachinery of suggested layout. - Abstract: The Supercritical Carbon Dioxide (S-CO{sub 2}) Brayton cycle has been gaining attention due to its compactness and high efficiency at moderate turbine inlet temperature. Previous S-CO{sub 2} cycle research works in the field of nuclear engineering were focused on its application to the next generation reactor with higher turbine inlet temperature than the existing conventional water-cooled nuclear power plants. However, it was shown in authors’ previous paper that the advantages of the S-CO{sub 2} Brayton cycle can be also further applied to the water-cooled Small Modular Reactor (SMR) with a success, since SMR requires minimal overall footprint while retaining high performance. One of the major issues in the S-CO{sub 2} Brayton cycle is the selection and design of appropriate turbomachinery for the designed cycle. Because most of the nuclear industry uses incompressible working fluids or ideal gases in the turbomachinery, a more detailed examination of the design of the turbomachinery is required for a power system that uses S-CO{sub 2} as working fluid. This is because the S-CO{sub 2} Brayton cycle high efficiency is the result of the non-ideal variation of properties near the CO{sub 2} critical point. Thus, the major focus of this paper is to suggest the design of the turbomachinery necessary for the S-CO{sub 2} Brayton cycle coupled to water cooled SMRs. For this reason, a S-CO{sub 2} Brayton cycle turbomachinery design methodology was suggested and the suggested design methodology was first tested with the existing experimental data to verify its capability. After then, it was

A charge regulating system for turbo-generator gas-cooled high-temperature reactor power stations

International Nuclear Information System (INIS)

Braytenbah, A.S.; Jaegtnes, K.O.

1975-01-01

The invention relates to a regulating system for gas-cooled high-temperature reactors power stations (helium coolant), equipped with several steam-boilers, each of which deriving heat from a corresponding cooling-gas flow circulating in the reactor, so as to feed superheated steam into a main common steam-manifold and re-superheated steam into a re-superheated hot common manifold [fr

Nuclear Power Station Kalkar, 300 MWe Prototype Nuclear Power Plant with Fast Sodium Cooled Reactor (SNR-300), Plant description

International Nuclear Information System (INIS)

1984-06-01

The nuclear power station Kalkar (SNR-300) is a prototype with a sodium cooled fast reactor and a thermal power of 762 MW. The present plant description has been made available in parallel to the licensing procedure for the reactor plant and its core Mark-Ia as supplementary information for the public. The report gives a detailed description of the whole plant including the prevention measures against the impact of external and plant internal events. The radioactive materials within the reactor cooling system and the irradiation protection and surveillance measures are outlined. Finally, the operation of the plant is described with the start-up procedures, power operation, shutdown phases with decay heat removal and handling procedures

Scientific-technical cooperation with Russia. Transient analyses for alternative types of water-cooled reactors. Final report

International Nuclear Information System (INIS)

Rohde, Ulrich; Pivovarov, Valeri; Matveev, Yurij

2010-12-01

The recently developed multi-group version DYN3D-MG of the reactor dynamics code DYN3D has been qualified for applications to water-cooled reactor concepts different from industrial PWR and BWR. An extended DYN3D version was applied to the graphite-moderated pressure tube reactor EGP-6 (NPP Bilibino) and conceptual design studies of an advanced Boiling Water Reactor with reduced moderation (RMWR) as well as the RUTA-70 reactor for low temperature heat supply. Concerning the RUTA reactor, safe heat removal by natural circulation of the coolant at low pressure has to be shown. For the corresponding validation of thermo-hydraulic system codes like ATHLET and RELAP5, experiments on flashing-induced natural circulation instabilities performed at the CIRCUS test facility at the TU Delft were simulated using the RELAP5 code. For the application to alternative water-cooled reactors, DYN3D model extensions and modifications were implemented, in particular adaptations of heat conduction and heat transfer models. Performing code-to-code comparisons with the Russian fine-mesh neutron diffusion code ACADEM contributed to the verification of DYN3D-MG. Validation has been performed by calculating reactor dynamics experiments at the NPP Bilibino. For the reactors EGP-6, RMWR and RUTA, analyses of various protected and unprotected control rod withdrawal and ejection transients were performed. The beyond design basis accident (BDBA) scenario ''Coast-down of all main coolant pumps at nominal power without scram'' for the RUTA reactor was analyzed using the code complexes DYN3D/ATHLET and DYN3D/RELAP5. It was shown, that the reactor passes over to a save asymptotic state at reduced power with coolant natural circulation. Analyzing the BDBA ''Unprotected withdrawal of a control rod group'' for the RMWR, the safety against Departure from Nucleate Boiling (DNB) could not be shown with the necessary confidence. Finally, conclusions have been drawn

Emergency reactor container cooling facility

International Nuclear Information System (INIS)

Suzuki, Hiroaki; Matsumoto, Tomoyuki.

1992-01-01

The present invention concerns an emergency cooling facility for a nuclear reactor container having a pressure suppression chamber, in which water in the suppression chamber is effectively used for cooling the reactor container. That is, the lower portion of a water pool in the pressure suppression chamber and the inside of the reactor container are connected by a pipeline. The lower end of the pipeline and a pressurized incombustible gas tank disposed to the outside of the reactor container are connected by a pipeline by way of valves. Then, when the temperature of the lower end of the pressure vessel exceeds a predetermined value, the valves are opened. If the valves are opened, the incombustible gas flows into the lower end of the pipeline connecting the lower portion of the water pool in the pressure suppression chamber and the inside of the reactor container. Since the inside of the pipeline is a two phase flow comprising a mixture of a gas phase and a liquid phase, the average density is decreased. Therefore, the water level of the two phase flow is risen by the level difference between the inside and the outside of the pipeline and, finally, the two phase mixture is released into the reactor container. As a result, the reactor container can be cooled by water in the suppression chamber by a static means without requiring pumps. (I.S.)

Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

International Nuclear Information System (INIS)

Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji

2000-01-01

Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

Energy Technology Data Exchange (ETDEWEB)

Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-06-01

Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production

Energy Technology Data Exchange (ETDEWEB)

Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

2002-01-01

The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

Steam generator tube failures: world experience in water-cooled nuclear power reactors in 1975

International Nuclear Information System (INIS)

Hare, M.G.

1976-11-01

Steam generator tube failures were reported in 22 out of 62 water-cooled nuclear power plants surveyed in 1975. This was less than in 1974, and the number of the tubes affected was noticeably less. This report summarizes these failures, most of which were due to corrosion. Secondary-water chemistry control, procedures for inspection and repair, tube materials, and failure rates are discussed. (author)

An alternative solution for heavy liquid metal cooled reactors fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Vitale Di Maio, Damiano, E-mail: damiano.vitaledimaio@uniroma1.it [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Cretara, Luca; Giannetti, Fabio [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Peluso, Vincenzo [“ENEA”, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Gandini, Augusto [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Manni, Fabio [“SRS Engineering Design S.r.l.”, Vicolo delle Palle 25-25/b, 00186 Rome (Italy); Caruso, Gianfranco [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy)

2014-10-15

Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution.

Heat removal performance of auxiliary cooling system for the high temperature engineering test reactor during scrams

International Nuclear Information System (INIS)

Takeda, Takeshi; Tachibana, Yukio; Iyoku, Tatsuo; Takenaka, Satsuki

2003-01-01

The auxiliary cooling system of the high temperature engineering test reactor (HTTR) is employed for heat removal as an engineered safety feature when the reactor scrams in an accident when forced circulation can cool the core. The HTTR is the first high temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 degree sign C and thermal power of 30 MW. The auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components and water boiling of itself. Simulation tests on manual trip from 9 MW operation and on loss of off-site electric power from 15 MW operation were carried out in the rise-to-power test up to 20 MW of the HTTR. Heat removal characteristics of the auxiliary cooling system were examined by the tests. Empirical correlations of overall heat transfer coefficients were acquired for a helium/water heat exchanger and air cooler for the auxiliary cooling system. Temperatures of fluids in the auxiliary cooling system were predicted on a scram event from 30 MW operation at 950 degree sign C of the reactor outlet coolant temperature. Under the predicted helium condition of the auxiliary cooling system, integrity of fuel blocks among the core graphite components was investigated by stress analysis. Evaluation results showed that overcooling to the core graphite components and boiling of water in the auxiliary cooling system should be prevented where open area condition of louvers in the air cooler is the full open

Emergency cooling system with hot-water jet pumps for nuclear reactors

International Nuclear Information System (INIS)

Reinsch, A.O.W.

1977-01-01

The ECCS for a PWR or BWR uses hot-water jet pumps to remove the thermal energy generated in the reactor vessel and stored in the water. The hot water expands in the nozzle part (Laval nozzle) of the jet pump and sucks in coolant (borated water) coming from a storage tank containing subcooled water. This water is mixing with the hot water/steam mixture from the Laval nozzle. The steam is condensed. The kinetic energy of the water is converted into a pressure increase which is sufficient to feed the water into the reactor vessel. The emergency cooling may further be helped by a jet condenser also operating according to the principle of a jet pump and condensing the steam generated in the reactor vessel. (DG) [de

Emergency cooling system for a gas-cooled nuclear reactor

International Nuclear Information System (INIS)

Cook, R.K.; Burylo, P.S.

1975-01-01

The site of the gas-cooled reactor with direct-circuit gas turbine is preferably the sea coast. An emergency cooling system with safety valve and emergency feed-water addition is designed which affects at least a part of the reactor core coolant after leaving the core. The emergency cooling system includes a water emergency cooling circuit with heat exchanger for the core coolant. The safety valve releases water or steam from the emergency coolant circuit when a certain temperature is exceeded; this is, however, replaced by the emergency feed-water. If the gas turbine exhibits a high and low pressure turbine stage, which are flowed through by coolant one behind another, a part of the coolant can be removed in front of each part turbine by two valves and be added to the haet exchanger. (RW/LH) [de

STUDY OF WATER HAMMERS IN THE FILLING OF THE SYSTEM OF PRESSURE COMPENSATION IN THE WATER-COOLED AND WATER-MODERATED POWER REACTORS

Directory of Open Access Journals (Sweden)

A. V. Korolyev

2017-01-01

list of initial events of severe accidents at NPPs with a water-cooled and water-moderated power reactor can be expanded.

Analysis of water cooled reactors stability; Analiza stabilnosti reaktorskih sistema hladjenih vodom

Energy Technology Data Exchange (ETDEWEB)

Marinkovic, P; Pesic, M [Boris Kidric Institute of Nuclear Sciences, Vinca, Belgrade (Yugoslavia)

1980-07-01

A model for stability analysis of non-boiling water cooled nuclear system is developed. The model is based on linear reactor kinetics and space averaged heat transfer in reactor and heat-exchanger. The transfer functions are defined and the analysis was applied to nuclear reactor RA at 'Boris Kidric' Institute - Vinca. (author)

Investigation of primary cooling water chemistry following the partial meltdown of Pu-Be neutron source in Tehran Research Reactor Core (TRR)

Energy Technology Data Exchange (ETDEWEB)

Aghoyeh, Reza Gholizadeh [School of Research and Development of Nuclear Reactors and Accelerators, Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran (AEOI), P.O. Box: 14155-1339, Tehran (Iran, Islamic Republic of); Khalafi, Hossein, E-mail: hkhalafi@aeoi.org.i [School of Research and Development of Nuclear Reactors and Accelerators, Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran (AEOI), P.O. Box: 14155-1339, Tehran (Iran, Islamic Republic of)

2011-03-15

Research highlights: Effect of Pu-Be neutron source meltdown in core on reactor water chemistry. Water chemistry of primary cooling before, during and after of above incident was compared. Training importance. Management of nuclear incident and accident. - Abstract: Effect of Pu-Be neutron source meltdown in core on reactor water chemistry was main aim of this study. Leaving the neutron source in the core after reactor power exceeds a few hundred Watts was the main reason for its partial meltdown. Water chemistry of primary cooling before, during and after of above incident was compared. Activity of some radio-nuclides such as Ba-140, La-140, I-131, I-132, Te-132 and Xe-135 increased. Other radio-nuclides such as Nd-147, Xe-133, Sr-91, I-133 and I-135 are also detected which were not existed before this incident.

Steam generators in indirect-cycle water-cooled reactors

International Nuclear Information System (INIS)

Fajeau, M.

1976-01-01

In the indirect cycle water-cooled nuclear reactors, the steam generators are placed between the primary circuit and the turbine. They act both as an energy transmitter and as a leaktigh barrier against fission or corrosion products. Their study is thus very important from a performance and reliability point of view. Two main types are presented here: the U-tube and the once-through steam generators [fr

Linear Dynamics Model for Steam Cooled Fast Power Reactors

Energy Technology Data Exchange (ETDEWEB)

Vollmer, H

1968-04-15

A linear analytical dynamic model is developed for steam cooled fast power reactors. All main components of such a plant are investigated on a general though relatively simple basis. The model is distributed in those parts concerning the core but lumped as to the external plant components. Coolant is considered as compressible and treated by the actual steam law. Combined use of analogue and digital computer seems most attractive.

Advanced gas-cooled reactors (AGR)

Energy Technology Data Exchange (ETDEWEB)

Yeomans, R. M. [South of Scotland Electricity Board, Hunterston Power Station, West Kilbride, Ayshire, UK

1981-01-15

The paper describes the advanced gas-cooled reactor system, Hunterston ''B'' power station, which is a development of the earlier natural uranium Magnox type reactor. Data of construction, capital cost, operating performance, reactor safety and also the list of future developments are given.

Gas cooled reactor assessment. Volume II. Final report, February 9, 1976--June 30, 1976

International Nuclear Information System (INIS)

1976-08-01

This report was prepared to document the estimated power plant capital and operating costs, and the safety and environmental assessments used in support of the Gas Cooled Reactor Assessment performed by Arthur D. Little, Inc. (ADL), for the U.S. Energy Research and Development Administration. The gas-cooled reactor technologies investigated include: the High Temperature Gas Reactor Steam Cycle (HTGR-SC), the HTGR Direct Cycle (HTGR-DC), the Very High Temperature Reactor (VHTR) and the Gas Cooled Fast Reactor (GCFR). Reference technologies used for comparison include: Light Water Reactors (LWR), the Liquid Metal Fast Breeder Reactor (LMFBR), conventional coal-fired steam plants, and coal combustion for process heat

Experimental tests and qualification of analytical methods to address thermohydraulic phenomena in advanced water cooled reactors. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

2000-05-01

Worldwide there is considerable experience in nuclear power technology, especially in water cooled reactor technology. Of the operating plants, in September 1998, 346 were light water reactors (LWRs) totalling 306 GW(e) and 29 were heavy water reactors (HWRs) totalling 15 GW(e). The accumulated experience and lessons learned from these plants are being incorporated into new advanced reactor designs. Utility requirements documents have been formulated to guide these design activities by incorporating this experience, and results from research and development programmes, with the aim of reducing costs and licensing uncertainties by establishing the technical bases for the new designs. Common goals for advanced designs are high availability, user-friendly features, competitive economics and compliance with internationally recognized safety objectives. Large water cooled reactors with power outputs of 1300 MW(e) and above, which possess inherent safety characteristics (e.g. negative Doppler moderator temperature coefficients, and negative moderator void coefficient) and incorporate proven, active engineered systems to accomplish safety functions are being developed. Other designs with power outputs from, for example, 220 MW(e) up to about 1300 MW(e) which also possess inherent safety characteristics and which place more emphasis on utilization of passive safety systems are being developed. Passive systems are based on natural forces and phenomena such as natural convection and gravity, making safety functions less dependent on active systems and components like pumps and diesel generators. In some cases, further experimental tests for the thermohydraulic conditions of interest in advanced designs can provide improved understanding of the phenomena. Further, analytical methods to predict reactor thermohydraulic behaviour can be qualified for use by comparison with the experimental results. These activities should ultimately result in more economical designs. The

Estimation, comparison, and evaluation of advanced fission power reactor generation costs

International Nuclear Information System (INIS)

Waddell, J.D.

1977-01-01

The study compares the high-temperature gas-cooled reactor (HTGR), the gas-cooled fast reactor (GCFR), the molten-salt breeder reactor (MSBR), the light water breeder reactor (LWBR), and the heavy water reactor (HWR) with proposed light water reactors (LWR) and liquid-metal fast breeder reactors (LMFBR). The relative electrical generation costs, including the effects of the introduction of advanced reactor fuel cycles into the U.S. nuclear power economy, were projected through the year 2030. The study utilized the NEEDS computer code which is a simulation of the U.S. nuclear power economy. The future potential electrical generation costs and cumulative consumption of uranium ore were developed using characterizations of the advanced systems. The reactor-fuel cycle characterizations were developed from literature reviews and personal discussions with the proponents of the various systems. The study developed a ranking of the concepts based on generation costs and uranium consumption

78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

Science.gov (United States)

2013-10-25

... comments were received. A companion guide, DG-1277, ``Initial Test Program of Emergency Core Cooling... NUCLEAR REGULATORY COMMISSION [NRC-2011-0129] Preoperational Testing of Emergency Core Cooling... (RG), 1.79, ``Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors...

Assessment of the thorium fuel cycle in power reactors

International Nuclear Information System (INIS)

Kasten, P.R.; Homan, F.J.; Allen, E.J.

1977-01-01

A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled

Plugging inaccessible leaks in cooling water pipework in nuclear power plants

International Nuclear Information System (INIS)

Powell, A.B.; May, R.; Down, M.G.

1988-01-01

The manifestation of initially small leaks in ancilliary reactor cooling water systems is not an unusual event. Often these leaks are in virtually inaccessible locations - for example, buried in thick concrete shielding or situated in cramped and highly radioactive vaults. Such leaks may ultimately prejudice the availability of the entire nuclear system. Continued operation without repair can result in the leak becoming larger, and the leaking water can cause further corrosion problems and interfere with instrumentation. In addition, the water may increase the volume of radwaste. In short, initially trivial leaks may cause significant operating problems. This paper describes the sealing of such leaks in the biological shield cooling system of Ontario Hydro's Pickering nuclear generating station CANDU reactors

Improving the understanding of thermal-hydraulics and heat transfer for super critical water cooled reactors

International Nuclear Information System (INIS)

Bilbao y Leon, S.; Aksan, N.

2010-01-01

Ensuring the exchange of information and fostering the collaboration among Member States on the development of technology advances for future nuclear power plants are among the key roles of the IAEA. There is high interest internationally in both developing and industrialized countries in the design of innovative super-critical water-cooled reactors (SCWRs). This interest arises from the high thermal efficiencies (44-45%) and improved economic competitiveness promised by for this concept, utilizing and building on the recent developments of highly efficient fossil power plants. The SCWR is one of the six concepts included in the Generation-IV International Forum (GIF). Following the advice of the IAEA Nuclear Energy Dept.'s Technical Working Groups on Advanced Technologies for LWRs and HWRs (the TWG-LWR and TWG-HWR), with the feedback from the Gen-IV SCWR Steering Committee, and in coordination with the OECD-NEA, IAEA is working on a Coordinated Research Project (CRP) in the areas of heat transfer behaviour and testing of thermo-hydraulic computer methods for Supercritical Water-Cooled Reactors. The second Research Coordination Meeting (RCM) of the CRP was held at the IAEA Headquarters, in Vienna (Austria)) in August 2009. This paper summarizes the current status of the CRP, as well as the major achievements to date. (authors)

Numerical simulation of severe water ingress accidents in a modular high temperature gas cooled reactor

International Nuclear Information System (INIS)

Zhang Zuoyi; Scherer, W.

1996-01-01

This report analyzes reverse water ingress accidents in the SIEMENS 200 MW Modular Pebble-Bed High Temperature Gas Cooled Reactor (HTR-MODULE) under the assumption of no active safety protection systems in order to find the safety margins of the current HTR-MODULE design and to realize a catastrophe-free nuclear technology. A water, steam and helium multi-phase cavity model is developed and implemented in the DSNP simulation system. The DSNP system is then used to simulate the primary and secondary circuit of a HTR-MODULE power plant. Comparisons of the model with experiments and with TINTE calculations serve as validation of the simulation. The analysis of the primary circuit tries to answer the question how fast the water enters the reactor core. It was found that the maximum H 2 O concentration increase in the reactor core is smaller than 0.3 kg/(m 3 s). The liquid water vaporization in the steam generator and H 2 O transport from the steam generator to the reactor core reduce the ingress velocity of the H 2 O into the reactor core. In order to answer the question how much water enters the primary circuit, the full cavitation of the feed water pumps is analyzed. It is found that if the secondary circuit is depressurized enough, the feed water pumps will be inherently stopped by the full cavitation. This limits the water to be pumped from the deaerator to the steam generator. A comprehensive simulation of the MODUL-HTR power plant then shows that the H 2 O inventory in the primary circuit can be limited to about 3000 kg. The nuclear reactivity increase caused by the water ingress leads to a fast power excursion, which, however, is inherently counterbalanced by negative feedback effects. Concerning the integrity of the fuel elements, the safety relevant temperature limit of 1600 C was not reached in any case. (orig.) [de

Cascade: a high-efficiency ICF power reactor

International Nuclear Information System (INIS)

Pitts, J.H.

1985-01-01

Cascade attains a net power-plant efficiency of 49% and its cost is competitive with high-temperature gas-cooled reactor, pressurized-water reactor, and coal-fired power plants. The Cascade reactor and blanket are made of ceramic materials and activation is 6 times less than that of the MARS Tandem Mirror Reactor operating at comparable power. Hands-on maintenance of the heat exchangers is possible one day after shutdown. Essentially all tritium is recovered in the vacuum system, with the remainder recovered from the helium power conversion loop. Tritium leakage external to the vacuum system and power conversion loop is only 0.03 Ci/d

Revision of the second basic plans of power reactor development in Power Reactor and Nuclear Fuel Development Corporation

International Nuclear Information System (INIS)

1978-01-01

Revision of the second basic plans concerning power reactor development in PNC (Power Reactor and Nuclear Fuel Development Corporation) is presented. (1) Fast breeder reactors: As for the experimental fast breeder reactor, after reaching the criticality, the power is raised to 50 MW thermal output within fiscal 1978. The prototype fast breeder reactor is intended for the electric output of 200 MW -- 300 MW, using mixed plutonium/uranium oxide fuel. Along the above lines, research and development will be carried out on reactor physics, sodium technology, machinery and parts, nuclear fuel, etc. (2) Advanced thermal reactor: The prototype advanced thermal reactor, with initial fuel primarily of slightly enriched uranium and heavy water moderation and boiling water cooling, of 165 MW electric output, is brought to its normal operation by the end of fiscal 1978. Along the above lines, research and development will be carried out on reactor physics, machinery and parts, nuclear fuel, etc. (Mori, K

Reactor technology: power conversion systems and reactor operation and maintenance

International Nuclear Information System (INIS)

Powell, J.R.

1977-01-01

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

Preliminary Design of KAIST Micro Modular Reactor with Dry Air Cooling

Energy Technology Data Exchange (ETDEWEB)

Baik, Seung Joon; Bae, Seong Jun; Kim, Seong Gu; Lee, Jeong Ik [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2014-05-15

KAIST research team recently proposed a Micro Modular Reactor (MMR) concept which integrates power conversion unit (PCU) with the reactor core in a single module. Using supercritical CO{sub 2} as a working fluid of cycle can achieve physically compact size due to small turbomachinery and heat exchangers. The objective of this project is to develop a concept that can operate at isolated area. The design focuses especially on the operation in the inland area where cooling water is insufficient. Thus, in this paper the potential for dry air cooling of the proposed reactor will be examined by sizing the cooling system with preliminary approach. The KAIST MMR is a recently proposed concept of futuristic SMR. The MMR size is being determined to be transportable with land transportation. Special attention is given to the MMR design on the dry cooling, which the cooling system does not depend on water. With appropriately designed air cooling heat exchanger, the MMR can operate autonomously. Two types of air cooling methods are suggested. One is using fan and the other is utilizing cooling tower for the air flow. With fan type air cooling method it consumes about 0.6% of generated electricity from the nuclear reactor. Cooling tower occupies an area of 227 m{sup 2} and 59.6 m in height. This design is just a preliminary estimation of the dry cooling method, and therefore more detailed and optimal design will be followed in the next phase.

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.)

Cooling nuclear reactor fuel

International Nuclear Information System (INIS)

Porter, W.H.L.

1975-01-01

Reference is made to water or water/steam cooled reactors of the fuel cluster type. In such reactors it is usual to mount the clusters in parallel spaced relationship so that coolant can pass freely between them, the coolant being passed axially from one end of the cluster in an upward direction through the cluster and being effective for cooling under normal circumstances. It has been suggested, however, that in addition to the main coolant flow an auxiliary coolant flow be provided so as to pass laterally into the cluster or be sprayed over the top of the cluster. This auxiliary supply may be continuously in use, or may be held in reserve for use in emergencies. Arrangements for providing this auxiliary cooling are described in detail. (U.K.)

Core Design and Deployment Strategy of Heavy Water Cooled Sustainable Thorium Reactor

Directory of Open Access Journals (Sweden)

Naoyuki Takaki

2012-08-01

Full Text Available Our previous studies on water cooled thorium breeder reactor based on matured pressurized water reactor (PWR plant technology concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array and using heavy water as coolant is appropriate for achieving better breeding performance and higher burn-up simultaneously [1–6]. One optimum core that produces 3.5 GW thermal energy using Th-²³³U oxide fuel shows a breeding ratio of 1.07 and averaged burn-up of about 80 GWd/t with long cycle length of 1300 days. The moderator to fuel volume ratio is 0.6 and required enrichment of ²³³U for the fresh fuel is about 7%. The coolant reactivity coefficient is negative during all cycles despite it being a large scale breeder reactor. In order to introduce this sustainable thorium reactor, three-step deployment scenario, with intermediate transition phase between current light water reactor (LWR phase and future sustainer phase, is proposed. Both in transition phase and sustainer phase, almost the same core design can be applicable only by changing fissile materials mixed with thorium from plutonium to ²³³U with slight modification in the fuel assembly design. Assuming total capacity of 60 GWe in current LWR phase and reprocessing capacity of 800 ton/y with further extensions to 1600 ton/y, all LWRs will be replaced by heavy water cooled thorium reactors within about one century then thorium reactors will be kept operational owing to its potential to sustain fissile fuels while reprocessing all spent fuels until exhaustion of massive thorium resource.

Exergy analysis of a system using a chemical heat pump to link a supercritical water-cooled nuclear reactor and a thermochemical water splitting cycle

International Nuclear Information System (INIS)

Granovskii, M.; Dincer, I.; Rosen, M. A.; Pioro, I

2007-01-01

The power generation efficiency of nuclear plants is mainly determined by the permissible temperatures and pressures of the nuclear reactor fuel and coolants. These parameters are limited by materials properties and corrosion rates and their effect on nuclear reactor safety. The advanced materials for the next generation of CANDU reactors, which employ steam as a coolant and heat carrier, permit the increased steam parameters (outlet temperature up to 625 degree C and pressure of about 25 MPa). Supercritical water-cooled (SCW) nuclear power plants are expected to increase the power generation efficiency from 35 to 45%. Supercritical water-cooled nuclear reactors can be linked to thermochemical water splitting cycles for hydrogen production. An increased steam temperature from the nuclear reactor makes it also possible to utilize its energy in thermochemical water splitting cycles. These cycles are considered by many as one of the most efficient ways to produce hydrogen from water and to have advantages over traditional low-temperature water electrolysis. However, even lower temperature water splitting cycles (Cu-Cl, UT-3, etc.) require a heat supply at the temperatures over 550-600 degree C. A sufficient increase in the heat transfer from the nuclear reactor to a thermochemical water splitting cycle, without jeopardizing nuclear reactor safety, might be effectively achieved by application of a heat pump which increases the temperature the heat supplied by virtue of a cyclic process driven by mechanical or electrical work. A high temperature chemical heat pump which employs the reversible catalytic methane conversion reaction is proposed. The reaction shift from exothermic to endothermic and back is achieved by a change of the steam concentration in the reaction mixture. This heat pump, coupled with a SCW nuclear plant on one side and thermochemical water splitting cycle on the other, increases the temperature level of the 'nuclear' heat and, thus, the intensity of

Containment atmosphere cooling system for experimental fast reactor 'JOYO'

International Nuclear Information System (INIS)

Sasaki, Mikio; Hoshi, Akio; Sato, Morihiko; Takeuchi, Kaoru

1979-01-01

The experimental fast reactor ''JOYO'', the first sodium-cooled fast reactor in Japan, achieved the initially licensed full power operation (50 MW) in July 1978 and is now under steady operation. Toshiba has participated in the construction of this reactor as a leading manufacturer and supplied various systems. This article outlines the design philosophy, system concepts and the operating experience of the containment atmosphere cooling system which has many design interfaces throughout the whole plant and requires especially high reliability. The successful performance of this system during the reactor full-power operation owes to the spot cooling design philosophy and to the preoperational adjustment of heat load during the preheating period of reactor cooling system peculiar to FBR. (author)

Method and apparatus for emergency cooling of a nuclear power plant

International Nuclear Information System (INIS)

Naito, Masanori; Chino, Koichi; Sato, Chikara; Inoue, Hisamichi.

1978-01-01

Purpose: To improve the cooling effect of spray water by eliminating the flow control effect for spray water due to increase in the steam pressure and flowing the entire spray water into the reactor core. Constitution: Upon emergency cooling of a reactor core by spraying coolants from above at the loss of coolant accident in a nuclear power plant, coolant is sprayed in a state where the temperature upon flowing into the reactor core is below the saturated temperature after heat exchange with vapors rising from the core. This enables to apply spray water always at a temperature and a flow rate in the range of whole volume falling irrespective of the water temperature in a pressure suppression pool. (Furukawa, Y.)

Outline of examination guides of water-cooled research reactors in Japan

International Nuclear Information System (INIS)

Yoshino, F.; Kimura, R.

1992-01-01

The Nuclear Safety Commission of Japan published two examination guides of water-cooled research reactors on July 18, 1991; one is for safety design, and another is for safety evaluation. In these guides, careful consideration is taken into account on the basic safety characteristic features of research reactors in order to be reasonable regulative requirements. This paper describes the fundamental philosophy and outline of the guides. (author)

Power generation costs for alternate reactor fuel cycles

International Nuclear Information System (INIS)

Smolen, G.R.; Delene, J.G.

1980-09-01

The total electric generating costs at the power plant busbar are estimated for various nuclear reactor fuel cycles which may be considered for power generation in the future. The reactor systems include pressurized water reactors (PWR), heavy-water reactors (HWR), high-temperature gas cooled reactors (HTGR), liquid-metal fast breeder reactors (LMFBR), light-water pre-breeder and breeder reactors (LWPR, LWBR), and a fast mixed spectrum reactor (FMSR). Fuel cycles include once-through, uranium-only recycle, and full recycle of the uranium and plutonium in the spent fuel assemblies. The U 3 O 8 price for economic transition from once-through LWR fuel cycles to both PWR recycle and LMFBR systems is estimated. Electric power generation costs were determined both for a reference set of unit cost parameters and for a range of uncertainty in these parameters. In addition, cost sensitivity parameters are provided so that independent estimations can be made for alternate cost assumptions

Fuel element replacement and cooling water activity at the musashi reactor

International Nuclear Information System (INIS)

Nozaki, Tetsuya; Honda, Teruyuki; Horiuchi, Norikazu; Aizawa, Otohiko; Sato, Tadashi

1989-01-01

The Musashi Institute of Technology Research Reactor (TRIGA 11, 100 kW) has been operated without serious problems since 1963. However, because there is no more spare fuel element, it was necessary to decide how to solve the problem. In the end, it was decided to obtain many stainless steel-clad fuel elements and operate with those fuel elements only, under the auspices of the Ministry of Education, Science and Culture. The bulk shielding experimental pool was remodeled as the storage for spent fuel elements, where the neutrons from the thermalizing column were shielded with cadmium and boron polyethylene plates. The equipment for transferring spent fuel elements was built and temporarily set up between the core tank and the new storage. These works were started in 1983, and finished in 1985. After the reactor was restarted, the count rate of the conventional cooling water monitor which was set in the cooling system using a GM counter drastically decreased. The spent fuel storage, the equipment and the works for fuel transfer, and the radioactivity of cooling water are reported. (K.I.)

Determining the void fraction in draught sections of a boiling water cooled reactor

International Nuclear Information System (INIS)

Fedulin, V.N.; Barolomej, G.G.; Solodkij, V.A.; Shmelev, V.E.

1987-01-01

Consideration is being given to the problem of improving methods for calculation of the void fraction in large channels of cooling system of the boiling water cooled reactor during two-phase unsteady flow. Investigation of the structure of two-phase flow was conducted in draught section of the VK-50 reactor (diameter D=2 m, height H=3). The method for calculation of the void fraction in channels with H/D ratio close to 1 is suggested

Development status and application prospect of supercritical-pressure light water cooled reactor

International Nuclear Information System (INIS)

Li Manchang; Wang Mingli

2006-01-01

The Supercritical-pressure Light Water Cooled Reactor (SCWR) is selected by the Generation IV International Forum (GIF) as one of the six Generation IV nuclear systems that will be developed in the future, and it is an innovative design based on the existing technologies used in LWR and supercritical coal-fired plants. Technically, SCWR may be based on the design, construction and operation experiences in existing PWR and supercritical coal-fired plants, which means that there is no insolvable technology difficulties. Since PWR technology will be adopted in the near term and medium term projects in China, and considering the sustainable development of the technology, it is an inevitable choice to research and develop the nuclear system of supercritical light water cooled reactor. (authors)

Water vapor as a perspective coolant for fast reactors

International Nuclear Information System (INIS)

Kalafati, D.D.; Petrov, S.I.

1978-01-01

Based on analysis of foreign projects of nuclear power plants with steam-cooled fast reactors, it is shown that low breeding ratio and large doubling time were caused by using nickel alloys, high vapor pressure and small volume heat release. The possibility is shown of obtaining doubling time in the necessary limits of T 2 =10-12 years when the above reasons for steam-cooled reactors are eliminated. Favourable combination of thermophysical and thermodynamic properties of water vapor makes it perspective coolant for power fast reactors

Plutonium Recycle Test Reactor (PRTR). Operating Experience and Supporting R and D, Its Application to Heavy-Water Power Reactor Design and Operation

Energy Technology Data Exchange (ETDEWEB)

Harty, H. [Battelle Memorial Institute, Pacific Northwest Laboratories, Richland, WA (United States)

1968-04-15

Convincing answers to questions about heavy-water, pressure-tube, power reactors, e.g. pressure-tube serviceability, heavy-water management problems, long-term behaviour of special pressure-tube reactor components, and unique operating maintenance problems (compared to light-water reactors) must be based on actual operating experience with that type of reactor. PRTR operating experience and supporting R and D studies, although not always simple extrapolations to power reactors, can be summarized in a context applicable to future heavy-water power reactors, as follows: 1. Pressure-tube life, in a practical case, need not be limited by creep, gross hydriding, corrosion, or mechanical damage. The possibility that growth of a defect (perhaps service-induced) to a size that is critical under certain operating conditions, remains a primary unknown in pressure- tube life extrapolations. A pressure-tube failure in PRTR (combined with gross release of fuel material) proved only slightly more inconvenient, time consuming, and damaging to the reactor proper, than occurred with a gross failure of a fuel element in PRTR. 2. Routine operating losses of heavy water appear tractable in heavy-water-cooled power reactors; losses from low-pressure systems can be insignificant over the life of a plant. Non-routine losses may prove to be the largest component of loss over the life of a plant. 3. The performance of special components in PRTR, e.g. the calandria and shields, has not deteriorated despite being subjected to non-standard operating conditions. The calandria now contains a light-water reflector with single barrier separation from the heavy-water moderator. The carbon steel shields (containing carbon steel shot) show no deterioration based on pressure drop measurements and piping activation immediately outside the shields. The helium pressurization system (for primary coolant pressurization) remains a high maintenance system, and cannot be recommended for power reactors, based

COOLING WATER ISSUES AND OPPORTUNITIES AT U.S. NUCLEAR POWER PLANTS

Energy Technology Data Exchange (ETDEWEB)

Gary Vine

2010-12-01

This report has been prepared for the Department of Energy, Office of Nuclear Energy (DOE-NE), for the purpose of providing a status report on the challenges and opportunities facing the U.S. commercial nuclear energy industry in the area of plant cooling water supply. The report was prompted in part by recent Second Circuit and Supreme Court decisions regarding cooling water system designs at existing thermo-electric power generating facilities in the U.S. (primarily fossil and nuclear plants). At issue in the courts have been Environmental Protection Agency regulations that define what constitutes “Best Technology Available” for intake structures that withdraw cooling water that is used to transfer and reject heat from the plant’s steam turbine via cooling water systems, while minimizing environmental impacts on aquatic life in nearby water bodies used to supply that cooling water. The report was also prompted by a growing recognition that cooling water availability and societal use conflicts are emerging as strategic energy and environmental issues, and that research and development (R&D) solutions to emerging water shortage issues are needed. In particular, cooling water availability is an important consideration in siting decisions for new nuclear power plants, and is an under-acknowledged issue in evaluating the pros and cons of retrofitting cooling towers at existing nuclear plants. Because of the significant ongoing research on water issues already being performed by industry, the national laboratories and other entities, this report relies heavily on ongoing work. In particular, this report has relied on collaboration with the Electric Power Research Institute (EPRI), including its recent work in the area of EPA regulations governing intake structures in thermoelectric cooling water systems.

COOLING WATER ISSUES AND OPPORTUNITIES AT U.S. NUCLEAR POWER PLANTS

International Nuclear Information System (INIS)

Vine, Gary

2010-01-01

This report has been prepared for the Department of Energy, Office of Nuclear Energy (DOE-NE), for the purpose of providing a status report on the challenges and opportunities facing the U.S. commercial nuclear energy industry in the area of plant cooling water supply. The report was prompted in part by recent Second Circuit and Supreme Court decisions regarding cooling water system designs at existing thermo-electric power generating facilities in the U.S. (primarily fossil and nuclear plants). At issue in the courts have been Environmental Protection Agency regulations that define what constitutes 'Best Technology Available' for intake structures that withdraw cooling water that is used to transfer and reject heat from the plant's steam turbine via cooling water systems, while minimizing environmental impacts on aquatic life in nearby water bodies used to supply that cooling water. The report was also prompted by a growing recognition that cooling water availability and societal use conflicts are emerging as strategic energy and environmental issues, and that research and development (R and D) solutions to emerging water shortage issues are needed. In particular, cooling water availability is an important consideration in siting decisions for new nuclear power plants, and is an under-acknowledged issue in evaluating the pros and cons of retrofitting cooling towers at existing nuclear plants. Because of the significant ongoing research on water issues already being performed by industry, the national laboratories and other entities, this report relies heavily on ongoing work. In particular, this report has relied on collaboration with the Electric Power Research Institute (EPRI), including its recent work in the area of EPA regulations governing intake structures in thermoelectric cooling water systems.

RCCS Experiments and Validation for High Temperature Gas-Cooled Reactor

International Nuclear Information System (INIS)

Chang Oh; Cliff Davis; Goon C. Park

2007-01-01

A reactor cavity cooling system (RCCS), an air-cooled helical coil RCCS unit immersed in the water pool, was proposed to overcome the disadvantages of the weak cooling ability of air-cooled RCCS and the complex structure of water-cooled RCCS for the high temperature gas-cooled reactor (HTGR). An experimental apparatus was constructed to investigate the various heat transfer phenomena in the water pool type RCCS, such as the natural convection of air inside the cavity, radiation in the cavity, the natural convection of water in the water pool and the forced convection of air in the cooling pipe. The RCCS experimental results were compared with published correlations. The CFX code was validated using data from the air-cooled portion of the RCCS. The RELAP5 code was validated using measured temperatures from the reactor vessel and cavity walls

Emergency reactor cooling circuit

International Nuclear Information System (INIS)

Araki, Hidefumi; Matsumoto, Tomoyuki; Kataoka, Yoshiyuki.

1994-01-01

Cooling water in a gravitationally dropping water reservoir is injected into a reactor pressure vessel passing through a pipeline upon occurrence of emergency. The pipeline is inclined downwardly having one end thereof being in communication with the pressure vessel. During normal operation, the cooling water in the upper portion of the inclined pipeline is heated by convection heat transfer from the communication portion with the pressure vessel. On the other hand, cooling water present at a position lower than the communication portion forms cooling water lumps. Accordingly, temperature stratification layers are formed in the inclined pipeline. Therefore, temperature rise of water in a vertical pipeline connected to the inclined pipeline is small. With such a constitution, the amount of heat lost from the pressure vessel by way of the water injection pipeline is reduced. Further, there is no worry that cooling water to be injected upon occurrence of emergency is boiled under reduced pressure in the injection pipeline to delay the depressurization of the pressure vessel. (I.N.)

Safety actuator of the Cabri reactor as a function of its power and cooling fluid flow rate

International Nuclear Information System (INIS)

Bertrand, Jean; Da Costa Vieira, David; Tattegrain, Alain

1969-04-01

This report present a device which is to provide a stop command to the Cabri reactor when the rate of its power to the cooling fluid rate reaches a value determined with respect to water temperature in the circuit. The stop command is delivered by an actuator which opens a relay contact when the power reaches a specific value. The authors present the device, its characteristics, and principle. They also present the different amplifier circuits, the input and output circuits (flow rate input, temperature input, and output circuit), the energy supply, and the various adjustments

IAEA coordinated research project on thermal-hydraulics of Supercritical Water-Cooled Reactors (SCWRs)

International Nuclear Information System (INIS)

Yamada, K.; Aksan, S. N.

2012-01-01

The Supercritical Water-Cooled Reactor (SCWR) is an innovative water-cooled reactor concept, which uses supercritical pressure water as reactor coolant. It has been attracting interest of many researchers in various countries mainly due to its benefits of high thermal efficiency and simple primary systems, resulting in low capital cost. The IAEA started in 2008 a Coordinated Research Project (CRP) on Thermal-Hydraulics of SCWRs as a forum to foster the exchange of technical information and international collaboration in research and development. This paper summarizes the activities and current status of the CRP, as well as major progress achieved to date. At present, 15 institutions closely collaborate in several tasks. Some organizations have been conducting thermal-hydraulics experiments and analysing the data, and others have been participating in code-to-test and/or code-to-code benchmark exercises. The expected outputs of the CRP are also discussed. Finally, the paper introduces several IAEA activities relating to or arising from the CRP. (authors)

International working group on gas-cooled reactors. Summary report

Energy Technology Data Exchange (ETDEWEB)

1981-01-15

The purpose of the meeting was to provide a forum for exchange of information on safety and licensing aspects for gas-cooled reactors in order to provide comprehensive review of the present status and of directions for future applications and development. Contributions were made concerning the operating experience of the Fort St. Vrain (FSV) HTGR Power Plant in the United States of America, the experimental power station Arbeitsgemeinschaft Versuchsreaktor (AVR) in the Federal Republic of Germany, and the CO/sub 2/-cooled reactors in the United Kingdom such as Hunterson B and Hinkley Point B. The experience gained at each of these reactors has proved the high safety potential of Gas-cooled Reactor Power Plants.

Investigation in justification of innovation supercritical water-cooled reactor - WWER-SCP

International Nuclear Information System (INIS)

Kirillov, P.L.; Baranaev, Yu.D.; Bogoslovskaya, G.P.; Glebov, A.P.; Grabezhnaya, V.A.; Kartashov, K.V.; Klushin, A.V.; Popov, V.V.

2014-01-01

State-of-the-art, gathered experience and development prospects of water-cooled reactors of next generation are considered. It is pointed out that development of SCWR is more attractive from the viewpoint of the basis principle of infrastructure - NPP adaptation without excessive investments. The results of experimental and calculational study of reactor installations on supercritical parameters (SCP) of water and freon are given. Consideration is given to the data on heat transfer at SCP of coolant, optimization of thermodynamic cycle, codes for thermohydraulic calculations, processes of heat and mass transfer at SCP, mass transfer and corrosion in SCP water, fuel elements and martials [ru

ULTRA SCWR+: Practical advanced water reactor concepts

International Nuclear Information System (INIS)

Duffey, Romney; Khartabil, Hussam; Kuran, Sermet; Zhou, Tracy; Pioro, Igor

2008-01-01

Modern thermal power plants now utilize supercritical steam cycles with thermal efficiencies of over 45%. Recent developments have lead to Ultra-SuperCritical (USC) systems, which adopt reheat turbines that can attain efficiencies of over 50%. Because these turbines are already developed, demonstrated and deployed worldwide, and use existing and traditional steam cycle technology, the simplest nuclear advance is to utilize these proven thermal cycle conditions by coupling this turbine type to a reactor. This development direction is fundamentally counter to the usual approach of adopting high-temperature gas-cooled (helium-cooled) reactor cycles, for which turbines have yet to be demonstrated on commercial scale unlike the supercritical steam turbines. The ULTRA (Ultra-supercritical Light water Thermal ReActor) SCWR+ concept adopts the fundamental design approach of matching a water and steam-cooled reactor to the ultra-supercritical steam cycle, adopting the existing and planned thermal power plant turbines. The HP and IP sections are fed with conditions of 25 MPa/625degC and 7 MPa/700degC, respectively, to achieve operating plant thermal efficiencies in excess of 50%, with a direct turbine cycle. By using such low-pressure reheated steam, this concept also adopts technology that was explored and used many years ago in existing water reactors, with the potential to produce large quantities of low cost heat, which can be used for other industrial and district processes. Pressure-Tube (PT) reactors are suitable for adoption of this design approach and, in addition, have other advantages that will significantly improve water-cooled reactor technology. These additional advantages include enhanced safety and improved resource utilization and proliferation resistance. This paper describes the PT-SCWR+ concept and its potential enhancements. (author)

Updated heat transfer correlations for supercritical water-cooled reactor applications

International Nuclear Information System (INIS)

Mokry, S.J.; Pioro, I.L.; Farah, A.; King, K.

2011-01-01

In support of the development of SuperCritical Water-cooled Reactors (SCWRs), research is currently being conducted for heat-transfer at supercritical conditions. Currently, there are no experimental datasets for heat transfer from power reactor fuel bundles to the fuel coolant (Water) available in open literature. Therefore, for preliminary calculations, heat-transfer correlations obtained with bare tube data can be used as a conservative approach. A large set of experimental data, for supercritical water was analyzed and an updated heat-transfer correlation for forced-convective heat-transfer, in the normal heat transfer regime, was developed. This experimental dataset was obtained within conditions similar to those for proposed SCWR concepts. Thus, this new correlation can be used for preliminary heat-transfer calculations in SCWR fuel channels. It has demonstrated a good fit for the analyzed dataset. Experiments with SuperCritical Water (SCW) are very expensive. Therefore, a number of experiments are performed in modeling fluids, such as carbon dioxide and refrigerants. However, there is no common opinion if SC modeling fluids' correlations can be applied to SCW and vice versa. Therefore, a correlation for supercritical carbon dioxide heat transfer was developed as a less expensive alternative to using supercritical water. The conducted analysis also meets the objective of improving our fundamental knowledge of the transport processes and handling of supercritical fluids. These correlations can be used for supercritical water heat exchangers linked to indirect-cycle concepts and the cogeneration of hydrogen, for future comparisons with other independent datasets, with bundle data, for the verification of computer codes for SCWR core thermalhydraulics and for the verification of scaling parameters between water and modeling fluids. (author)

Numerical modeling of a nuclear production reactor cooling lake

International Nuclear Information System (INIS)

Hamm, L.L.; Pepper, D.W.

1987-01-01

A finite element model has been developed which predicts flow and temperature distributions within a nuclear reactor cooling lake at the Savannah River Plant near Aiken, South Carolina. Numerical results agree with values obtained from a 3-D EPA numerical lake model and actual measurements obtained from the lake. Because the effluent water from the reactor heat exchangers discharges directly into the lake, downstream temperatures at mid-lake could exceed the South Carolina DHEC guidelines for thermal exchanges during the summer months. Therefore, reactor power was reduced to maintain temperature compliance at mid-lake. Thermal mitigation measures were studied that included placing a 6.1 m deep fabric curtain across mid-lake and moving the reactor outfall upstream. These measurements were calculated to permit about an 8% improvement in reactor power during summer operation

Reactor cooling system

International Nuclear Information System (INIS)

Kato, Etsuji.

1979-01-01

Purpose: To eliminate cleaning steps in the pipelines upon reactor shut-down by connecting a filtrating and desalting device to the cooling system to thereby always clean up the water in the pipelines. Constitution: A filtrating and desalting device is connected to the pipelines in the cooling system by way of drain valves and a check valve. Desalted water is taken out from the exit of the filtrating and desalting device and injected to one end of the cooling system pipelines by way of the drain valve and the check valve and then returned by way of another drain valve to the desalting device. Water in the pipelines is thus always desalted and the cleaning step in the pipelines is no more required in the shut-down. (Kawakami, Y.)

Neutronic of heterogenous gas cooled reactors

International Nuclear Information System (INIS)

Maturana, Roberto Hernan

2008-01-01

At present, one of the main technical features of the advanced gas cooled reactor under development is its fuel element concept, which implies a neutronic homogeneous design, thus requiring higher enrichment compared with present commercial nuclear power plants.In this work a neutronic heterogeneous gas cooled reactor design is analyzed by studying the neutronic design of the Advanced Gas cooled Reactor (AGR), a low enrichment, gas cooled and graphite moderated nuclear power plant.A search of merit figures (some neutronic parameter, characteristic dimension, or a mixture of both) which are important and have been optimized during the reactor design stage is been done, to aim to comprise how a gas heterogeneous reactor is been design, given that semi-infinity arrangement criteria of rods in LWRs and clusters in HWRs can t be applied for a solid moderator and a gas refrigerator.The WIMS code for neutronic cell calculations is been utilized to model the AGR fuel cell and to calculate neutronic parameters such as the multiplication factor and the pick factor, as function of the fuel burnup.Also calculation is been done for various nucleus characteristic dimensions values (fuel pin radius, fuel channel pitch) and neutronic parameters (such as fuel enrichment), around the design established parameters values.A fuel cycle cost analysis is carried out according to the reactor in study, and the enrichment effect over it is been studied.Finally, a thermal stability analysis is been done, in subcritical condition and at power level, to study this reactor characteristic reactivity coefficients.Present results shows (considering the approximation used) a first set of neutronic design figures of merit consistent with the AGR design. [es

Materials challenges for the supercritical water-cooled reactor (SCWR)

International Nuclear Information System (INIS)

Baindur, S.

2008-01-01

This paper discusses the materials requirements of the Supercritical Water-cooled Reactor (SCWR) which arise from its severe expected operating conditions: (i) Outlet Temperature (to 650 C); (ii) Pressure of 25 MPa for the coolant containment, (iii) Thermochemical stress in the presence of supercritical water, and (iv) Radiative damage (up to 150 dpa for the fast spectrum variant). These operating conditions are reviewed; the phenomenology of materials in the supercritical water environment that create the materials challenges is discussed; knowledge gaps are identified, and efforts to understand material behaviour under the operating conditions expected in the SCWR are described. (author)

The discussion of nuclear power plant's cooling chain design for freezing site

International Nuclear Information System (INIS)

Hu Jian; Yang Ting; Jiang Xulun

2014-01-01

The Component cooling water system (RRI) and Essential service water system (SEC) are composed of Nuclear Power Plant's (NPP) cooling chain, which has its special requirement for freezing site from system design and safety point of view. The feature and difficulty of cooling chain design at freezing condition (when the intake water temperature is below O ℃) are represented. At present, several NPPs are in operation or under construction at freezing site in the world, including Pressurized Water Reactor (PWR) and Canadian Deuterium Uranium reactor (CANDU). By analyzing the thoughts and applicability of different kinds of cooling chain design at freezing site, one solution called 'SEC thermal discharge reflux' is proposed to remove the residual heat from Nuclear Island (NI) into heat sink safely in winter. The solution has been approved by National Nuclear Safety Administration (NNSA) in China and applied in one of CPR NPP in the north of China, which is able to solve several problems compared with the traditional solutions, such as 'Reactor low power operation', 'Reactor start-up for the first time', and 'Changeover of RRI/SEC trains in winter'. The solution is also able to prevent RRI/SEC heat exchanger from icing and avoid low flowrate in SEC pipes. Besides, considering of the economical efficiency, simple operation and control strategy is designed. (authors)

A preliminary definition of the parameters of an experimental natural - uranium, graphite - moderated, helium - cooled power reactor

International Nuclear Information System (INIS)

Baltazar, O.

1978-01-01

A preliminary study of the technical characteristic of an experiment at 32 MWe power with a natural uconium, graphite-moderated, helium cooled reactor is described. The national participation and the use of reactor as an instrument for the technological development of future high temperature gas cooled reactor is considered in the choice of the reactor type. Considerations about nuclear power plants components based in extensive bibliography about similar english GCR reactor is presented. The main thermal, neutronic an static characteristic and in core management of the nuclear fuel is stablished. A simplified scheme of the secondary system and its thermodynamic performance is determined. A scheme of parameters calculation of the reactor type is defined based in the present capacity of calculation developed by Coordenadoria de Engenharia Nuclear and Centro de Processamento de Dados, IEA, Brazil [pt

Reactor-core isolation cooling system with dedicated generator

International Nuclear Information System (INIS)

Nazareno, E.V.; Dillmann, C.W.

1992-01-01

This patent describes a nuclear reactor complex. It comprises a dual-phase nuclear reactor; a main turbine for converting phase-conversion energy stored by vapor into mechanical energy for driving a generator; a main generator for converting the mechanical energy into electricity; a fluid reservoir external to the reactor; a reactor core isolation cooling system with several components at least some of which require electrical power. It also comprises an auxiliary pump for pumping fluid from the reservoir into the reactor pressure vessel; an auxiliary turbine for driving the pump; control means for regulating the rotation rate of the auxiliary turbine; cooling means for cooling the control means; and an auxiliary generator coupled to the auxiliary turbine for providing at least a portion of the electrical power required by the components during a blackout condition

Advanced technologies for water cooled reactors 1990. Pt. 1

International Nuclear Information System (INIS)

1991-05-01

The meeting was attended by 20 participants from 12 countries who reviewed and discussed the status and progress of national programmes on advanced water-cooled reactors and recommended to the Scientific Secretary a comprehensive programme for 1991/1992 which would support technology development programmes in IWGATWR Member States. This summary report outlines the activities of IWGATWR since its Second Meeting in June 1988 and main results of the Third Meeting

XHM-1 alloy as a promising structural material for water-cooled fusion reactor components

International Nuclear Information System (INIS)

Solonin, M.I.; Alekseev, A.B.; Kazennov, Yu.I.; Khramtsov, V.F.; Kondrat'ev, V.P.; Krasina, T.A.; Rechitsky, V.N.; Stepankov, V.N.; Votinov, S.N.

1996-01-01

Experience gained in utilizing austenitic stainless steel components in water-cooled power reactors indicates that the main cause of their failure is the steel's propensity for corrosion cracking. In search of a material immune to this type of corrosion, different types of austenitic steels and chromium-nickel alloys were investigated and tested at VNIINM. This paper presents the results of studying physical and mechanical properties, irradiation and corrosion resistance in a water coolant at <350 C of the alloy XHM-1 as compared with austenitic stainless steels 00Cr16Ni15Mo3Nb, 00Cr20Ni25Nb and alloy 00Cr20Ni40Mo5Nb. Analysis of the results shows that, as distinct from the stainless steels studied, the XHM-1 alloy is completely immune to corrosion cracking (CC). Not a single induced damage was encountered within 50 to 350 C in water containing different amounts of chlorides and oxygen under tensile stresses up to the yield strength of the material. One more distinctive feature of the alloy compared to steels is that no change in the strength or total elongation is encountered in the alloy specimens irradiated to 32 dpa at 350 C. The XHM-1 alloy has adequate fabricability and high weldability characteristics. As far as its properties are concerned, the XHM-1 alloy is very promising as a material for water-cooled fusion reactor components. (orig.)

Power control device for heavy water moderated reactor

International Nuclear Information System (INIS)

Matsushima, Hidesuke; Masuda, Hiroyuki.

1978-01-01

Purpose: To improve self controllability of a nuclear power plant, as well as enable continuous power level control by a controlled flow of moderators in void pipes provided in a reactor core. Constitution: Hollow void pipes are provided in a reactor core to which a heavy water recycle loop for power control, a heavy water recycle pump for power control, a heavy water temperature regulator and a heavy water flow rate control valve for power control are connected in series to constitute a heavy water recycle loop for flowing heavy water moderators. The void ratio in each of the void pipes are calculated by a process computer to determine the flow rate and the temperature for the recycled heavy water. Based on the above calculation result, the heavy water temperature regulator is actuated by way of a temperature setter at the heavy water inlet and the heavy water flow rate is controlled by the actuation of the heavy water flow rate control valve. (Kawakami, Y.)

Dual pressurized light water reactor producing 2000 M We

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-10-15

The dual unit optimizer 2000 M We (Duo2000) is proposed as a new design concept for large nuclear power plant. Duo is being designed to meet economic and safety challenges facing the 21 century green and sustainable energy industry. Duo2000 has two nuclear steam supply systems (NSSS) of the unit nuclear optimizer (Uno) pressurized water reactor (PWR) in a single containment so as to double the capacity of the plant. Uno is anchored to the optimized power reactor 1000 M We (OPR1000) of the Korea Hydro and Nuclear Power Co., Ltd. The concept of Duo can be extended to any number of PWRs or pressurized heavy water reactors (PHWR s), or even boiling water reactor (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. In particular, since it is required that the small and medium sized reactors (SMRs) be built as units, the concept of Duo2000 will apply to SMRs as well. With its in-vessel retention as severe accident management strategy, Duo can not only put the single most querulous PWR safety issue to end, but also pave ways to most promising large power capacity dispensing with huge redesigning cost for generation III + nuclear systems. The strengths of Duo2000 include reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting NSSS. The technology can further be extended to coupling modular reactors as dual, triple, or quadruple units to increase their economics, thus accelerating the commercialization as well as the customization of SMRs. (Author)

Study on Reactor Performance of Online Power Monitoring in PUSPATI TRIGA Reactor (RTP)

International Nuclear Information System (INIS)

Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

2014-01-01

The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on reactor performance of online power monitoring based on various parameter of reactor such as log power, linear power, period, Fuel and coolant temperature and reactivity parameter with using neutronic and other instrumentation system of reactor. Methodology of online power estimation and monitoring is to evaluate and analysis of reactor power which is important of reactor safety and control. Neutronic instrumentation system will use to estimate power measurement, differential of log and linear power and period during reactor operation .This study also focus on noise fluctuation from fission chamber during reactor operation .This work will present result of online power monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that optimization of online power monitoring will improved the reactor control and safety parameter of reactor during operation. (author)

The Effect of Topaz Irradiation to the Quality of Cooling Water Reactor GA Siwabessy

International Nuclear Information System (INIS)

Elisabeth Ratnawati; Kawkab Mustofa; Arif Hidayat

2012-01-01

Topaz irradiation which applied both inside and outside the reactor core is one utilization of the reactor GA Siwabessy. Topaz consists of silicon clusters containing a combination of aluminum, fluorine and hydroxyl, and impurities. The results of the qualitative analysis of the topaz before irradiation detected europium (Eu-152), potassium (K-40) and sodium (Na-24). While the post-irradiation of topaz detected europium (Eu), cobalt (Co), cesium (Cs), tantalum (Ta), scandium (Sc), iron (Fe), Selenium (Se) and potassium (K). These elements might affect the quality of the cooling water. But the results of the qualitative analysis that were carried out to the primary cooling water did not reveal any elements similar to the elements contained in topaz impurities. Most likely this is because most impurities have been caught by the resin trap in purification systems, because of the results of the analysis of the dirt on the resin trap contained elements similar to the impurities Fe and Co topaz. The purification system makes quality primary cooling water is maintained. From the result shows that chemically the quality of primary cooling water is not affected by the topaz irradiation. (author)

Proposal for Dual Pressurized Light Water Reactor Unit Producing 2000 MWe

International Nuclear Information System (INIS)

Kang, Kyoung Min; Noh, Sang Woo; Suh, Kune Yull

2009-01-01

The Dual Unit Optimizer 2000 MWe (DUO2000) is put forward as a new design concept for large power nuclear plants to cope with economic and safety challenges facing the 21 st century green and sustainable energy industry. DUO2000 is home to two nuclear steam supply systems (NSSSs) of the Optimized Power Reactor 1000 MWe (OPR1000)-like pressurized water reactor (PWR) in single containment so as to double the capacity of the plant. The idea behind DUO may as well be extended to combining any number of NSSSs of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactors (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to an end, but also pave the way to very promising large power capacity while dispensing with the huge redesigning cost for Generation III+ nuclear systems. Five prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The latent threats are discussed as well

International symposium on evolutionary water cooled reactors: strategic issues, technologies and economic viability. Book of extended synopses

International Nuclear Information System (INIS)

1998-01-01

Within the frame of growing energy demand caused by global economic growth and taking into account the Kyoto protocol on carbon dioxide emissions nuclear power plants attaining a new role. The presented papers deal mostly with improvements in NPP design, construction and safety. Some new concepts are proposed, especially in the field of inherent or passive reactor safety as well as computerised control systems. Water cooled reactors achieved already the necessary cost reduction but require some radical thinking in fuel design, construction rate, built-in safety. The key factor will be mass production in order to attain capital cost of half today's level

CEA programme on gas cooled reactors

International Nuclear Information System (INIS)

Carre, F.; Fiorini, G.L.; Chapelot, Ph.; Gauthier, J.C.

2002-01-01

Future nuclear energy systems studies conducted by the CEA aim at investigating and developing promising technologies for future reactors, fuels and fuel cycles, for nuclear power to play a major part in sustainable energy policies. Reactors and fuel cycles are considered as integral parts of a nuclear system to be optimised as a whole. Major goals assigned to future nuclear energy systems are the following: reinforced economic competitiveness with other electricity generation means, with a special emphasis on reducing the investment cost; enhanced reliability and safety, through an improved management of reactor operation in normal and abnormal plant conditions; minimum production of long lived radioactive waste; resource saving through an effective and flexible use of the available resources of fissile and fertile materials; enhanced resistance to proliferation risks. The three latter goals are essential for the sustainability of nuclear energy in the long term. Additional considerations such as the potentialities for other applications than electricity generation (co-generation, production of hydrogen, sea water desalination) take on an increasing importance. Sustainability goals call for fast neutron spectra (to transmute nuclear waste and to breed fertile fuel) and for recycling actinides from the spent fuel (plutonium and minor actinides). New applications and economic competitiveness call for high temperature technologies (850 deg C), that afford high conversion efficiencies and hence less radioactive waste production and discharged heat. These orientations call for breakthroughs beyond light water reactors. Therefore, as a result of a screening review of candidate technologies, the CEA has selected an innovative concept of high temperature gas cooled reactor with a fast neutron spectrum, robust refractory fuel, direct conversion with a gas turbine, and integrated on-site fuel cycle as a promising system for a sustainable energy development. This objective

Performance comparison of metallic, actinide burning fuel in lead-bismuth and sodium cooled fast reactors

International Nuclear Information System (INIS)

Weaver, K.D.; Herring, J.S.; Macdonald, P.E.

2001-01-01

Various methods have been proposed to ''incinerate'' or ''transmute'' the current inventory of transuranic waste (TRU) that exits in spent light-water-reactor (LWR) fuel, and weapons plutonium. These methods include both critical (e.g., fast reactors) and non-critical (e.g., accelerator transmutation) systems. The work discussed here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead and lead-alloy cooled fast reactors for producing low-cost electricity as well as for actinide burning. The neutronics of non fertile fuel loaded with 20 or 30-wt% light water reactor (LWR) plutonium plus minor actinides for use in a lead-bismuth cooled fast reactor are discussed in this paper, with an emphasis on the fuel cycle life and isotopic content. Calculations show that the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from -1.02 to -1.16 g/MWd, compared to a typical LWR actinide generation rate of 0.303 g/MWd. However, when using the same parameters, the sodium-cooled case went subcritical after 0.2 to 0.8 effective full power years, and the lead-bismuth cooled case ranged from 1.5 to 4.5 effective full power years. (author)

Recolonization of reactor cooling water system by the Asiatic clam Corbicula fluminea

International Nuclear Information System (INIS)

Harvey, R.S.

1978-01-01

Recolonization rates for the Asiatic clam Corbicula fluminea ranged from 3.0 to 5.6 metric tons per year in cooling water basins for a nuclear production reactor at the Savannah River Plant. However, a 10-month cleaning cycle for each basin (flow area, 6100 m 2 ) keeps the depth of the silt/clam layer low. With this cleaning frequency, Corbicula are not reaching heat exchangers at sufficient size or in sufficient numbers to restrict flow. Data are presented on the size/age distribution for clams recolonizing cooling water basins between cleanings

Emergency cooling system for a liquid metal cooled reactor

International Nuclear Information System (INIS)

Murata, Ryoichi; Fujiwara, Toshikatsu.

1980-01-01

Purpose: To suitably cool liquid metal as coolant in emergency in a liquid metal cooled reactor by providing a detector for the pressure loss of the liquid metal passing through a cooling device in a loop in which the liquid metal is flowed and communicating the detector with a coolant flow regulator. Constitution: A nuclear reactor is stopped in nuclear reaction by control element or the like in emergency. If decay heat is continuously generated for a while and secondary coolant is insufficiently cooled with water or steam flowed through a steam and water loop, a cooler is started. That is, low temperature air is supplied by a blower through an inlet damper to the cooler to cool the secondary coolant flowed into the cooler through a bypass pipe so as to finally safely stop an entire plant. Since the liquid metal is altered in its physical properties by the temperature at this time, it is detected to regulate the opening of the valve of the damper according to the detected value. (Sekiya, K.)

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.)

General design and main problems of a gas-heavy-water power reactor contained in a pressure vessel

International Nuclear Information System (INIS)

Roche, R.; Gaudez, J.C.

1964-01-01

In the framework of research carried out on a CO 2 -cooled power reactor moderated by heavy water, the so-called 'pressure vessel' solution involves the total integration of the core, of the primary circuit (exchanges and blowers) and of the fuel handling machine inside a single, strong, sealed vessel made of pre-stressed concrete. A vertical design has been chosen: the handling 'attic' is placed above the core, the exchanges being underneath. This solution makes it possible to standardize the type of reactor which is moderated by heavy-water or graphite and cooled by a downward stream of carbon dioxide gas; it has certain advantages and disadvantages with respect to the pressure tube solution and these are considered in detail in this report. Extrapolation presents in particular.problems due specifically to the heavy water (for example its cooling,its purification, the balancing of the pressures of the heavy water and of the gas, the assembling of the internal structures, the height of the attic, etc. (authors) [fr

Analysis on blow-down transient in water ingress accident of high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Wang, Yan; Zheng, Yanhua; Li, Fu; Shi, Lei

2014-01-01

Water ingress into the primary circuit is generally recognized as one of the severe accidents with potential hazard to the modular high temperature gas-cooled reactor, which will cause a positive reactivity introduction with the increase of steam density in reactor core to enhance neutron slowing-down, also the chemical corrosion of graphite fuel elements and the damage of reflector structure material. The increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes and flammable water gas. The research on water ingress transient is significant for the verification of inherent safety characteristics of high temperature gas-cooled reactor. The 200 MWe high temperature gas-cooled reactor (HTR-PM), designed by the Institute of Nuclear and New Energy Technology of Tsinghua University, is exampled to be analyzed in this paper. The design basis accident (DBA) scenarios of double-ended guillotine break of single heat-exchange tube (steam generator heat-exchange tube rupture) are simulated by the thermal-hydraulic analysis code, and some key concerns which are relative to the amount of water into the reactor core during the blow-down transient are analyzed in detail. The results show that both of water mass and steam ratio of the fluid spouting from the broken heat-exchange tube are affected by break location, which will increase obviously with the broken location closing to the outlet of the heat-exchange tube. The double-ended guillotine rupture at the outlet of the heat-exchange will result more steam penetrates into the reactor core in the design basis accident of water ingress. The mass of water ingress will also be affected by the draining system. It is concluded that, with reasonable optimization on design to balance safety and economy, the total mass of water ingress into the primary circuit of reactor could be limited effectively to meet the safety requirements, and the pollution of

Preliminary design of the cooling system for a gas-cooled, high-fluence fast pulsed reactor (HFFPR)

International Nuclear Information System (INIS)

Monteith, H.C.

1978-10-01

The High-Fluence Fast Pulsed Reactor (HFFPR) is a research reactor concept currently being evaluated as a source for weapon effects experimentation and advanced reactor safety experiments. One of the designs under consideration is a gas-cooled design for testing large-scale weapon hardware or large bundles of full-length, fast reactor fuel pins. This report describes a conceptual cooling system design for such a reactor. The primary coolant would be helium and the secondary coolant would be water. The size of the helium-to-water heat exchanger and the water-to-water heat exchanger will be on the order of 0.9 metre (3 feet) in diameter and 3 metres (10 feet) in length. Analysis indicates that the entire cooling system will easily fit into the existing Sandia Engineering Reactor Facility (SERF) building. The alloy Incoloy 800H appears to be the best candidate for the tube material in the helium-to-water heat exchanger. Type 316 stainless steel has been recommended for the shell of this heat exchanger. Estimates place the cost of the helium-to-water heat exchanger at approximately $100,000, the water-to-water heat exchanger at approximately $25,000, and the helium pump at approximately $450,000. The overall cost of the cooling system will approach $2 million

Cooling water distribution system

Science.gov (United States)

Orr, Richard

1994-01-01

A passive containment cooling system for a nuclear reactor containment vessel. Disclosed is a cooling water distribution system for introducing cooling water by gravity uniformly over the outer surface of a steel containment vessel using an interconnected series of radial guide elements, a plurality of circumferential collector elements and collector boxes to collect and feed the cooling water into distribution channels extending along the curved surface of the steel containment vessel. The cooling water is uniformly distributed over the curved surface by a plurality of weirs in the distribution channels.

Study of risk reduction by improving operation of reactor core isolation cooling system

International Nuclear Information System (INIS)

Watanabe, Yamato; Tazai, Ayuko; Yamagishi, Shohei; Muramatsu, Ken; Muta, Hitoshi

2014-01-01

The Fukushima Daiichi nuclear power plant fell into a station blackout (SBO) due to the earthquake and tsunami in which most of the core cooling systems were disabled. In the units 2 and 3, water injection to the core was performed only by water injection system with turbine driven pumps. In particular, it is inferred from observed plant parameters that the reactor core isolation cooling system (RCIC) continued its operation much longer than it was originally expected (8 hours). Since the preparation of safety measures did not work, the reactor core damaged. With a view to reduce risk of station blackout events in a BWR by accident management, this study investigated the efficacy of operation procedures that takes advantage of RCIC which can be operated with only equipment inside reactor building and does not require an AC power source. The efficacy was assessed in this study by two steps. The first step is a thermal hydraulic analysis with the RETRAN3D code to estimate the potential extension of duration of core cooling by RCIC and the second step is the estimation of time required for recovery of off-site power from experiences at nuclear power stations under the 3.11 earthquake. This study showed that it is possible to implement more reliable measures for accident termination and to greatly reduce the risk of SBO by the installation of accident management measures with use of RCIC for extension of core cooling under SBO conditions. (author)

Unconventional liquid metal cooled fast reactors

International Nuclear Information System (INIS)

Spinrad, B.I.; Rohach, A.F.; Razzaque, M.M.

1989-06-01

This report describes the rationale for, design of and analytical studies on an unconventional sodium-cooled power reactor, called the Trench Reactor. It derives its name from the long, narrow sodium pool in which the reactor is placed. Unconventional features include: pool shape; reactor shape (also long and narrow); reflector control; low power density; hot-leg primary pumping; absence of a cold sodium pool; large core boxes rather than a large number of subassemblies; large diameter metal fuel; vessel suspension from cables; and vessel cooling by natural circulation of building atmosphere (nitrogen) at all times. These features all seem feasible. They result in a system that is capable of at least a ten year reload interval and shows good safety through direct physical response to loss-of-heat-sink, loss-of-flow and limited-reactivity nuclear transients. 43 figs., 43 tabs

Advanced technologies for water cooled reactors 1990. Pt. 2

International Nuclear Information System (INIS)

1991-05-01

The main purpose of the meeting was to review and discuss the status of national programmes, the progress achieved since the last meeting held in June 1988 in the field of advanced technologies and design trends for existing and future water cooled reactors. 24 specialists from 14 countries and the IAEA took part in the meeting and 12 papers were presented. A separate abstract was prepared for each of these papers. Refs, figs and tabs

Design study of ship based nuclear power reactor

International Nuclear Information System (INIS)

Su'ud, Zaki; Fitriyani, Dian

2002-01-01

Preliminary design study of ship based nuclear power reactors has been performed. In this study the results of thermohydraulics analysis is presented especially related to behaviour of ship motion in the sea. The reactors are basically lead-bismuth cooled fast power reactors using nitride fuels to enhance neutronics and safety performance. Some design modification are performed for feasibility of operation under sea wave movement. The system use loop type with relatively large coolant pipe above reactor core. The reactors does not use IHX, so that the heat from primary coolant system directly transferred to water-steam loop through steam generator. The reactors are capable to be operated in difference power level during night and noon. The reactors however can also be used totally or partially to produce clean water through desalination of sea water. Due to the influence of sea wave movement the analysis have to be performed in three dimensional analysis. The computation time for this analysis is speeded up using Parallel Virtual Machine (PVM) Based multi processor system

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

International Nuclear Information System (INIS)

Brand, P.; Wiechers, W.K.

1974-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2003-04-01

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

Wetland Water Cooling Partnership: The Use of Constructed Wetlands to Enhance Thermoelectric Power Plant Cooling and Mitigate the Demand of Surface Water Use

Energy Technology Data Exchange (ETDEWEB)

Apfelbaum, Steven L. [Applied Ecological Services Inc., Brodhead, WI (United States); Duvall, Kenneth W. [Sterling Energy Services, LLC, Atlanta, GA (United States); Nelson, Theresa M. [Applied Ecological Services Inc., Brodhead, WI (United States); Mensing, Douglas M. [Applied Ecological Services Inc., Brodhead, WI (United States); Bengtson, Harlan H. [Sterling Energy Services, LLC, Atlanta, GA (United States); Eppich, John [Waterflow Consultants, Champaign, IL (United States); Penhallegon, Clayton [Sterling Energy Services, LLC, Atlanta, GA (United States); Thompson, Ry L. [Applied Ecological Services Inc., Brodhead, WI (United States)

2013-12-01

Through the Phase I study segment of contract #DE-NT0006644 with the U.S. Department of Energy’s National Energy Technology Laboratory, Applied Ecological Services, Inc. and Sterling Energy Services, LLC (the AES/SES Team) explored the use of constructed wetlands to help address stresses on surface water and groundwater resources from thermoelectric power plant cooling and makeup water requirements. The project objectives were crafted to explore and develop implementable water conservation and cooling strategies using constructed wetlands (not existing, naturally occurring wetlands), with the goal of determining if this strategy has the potential to reduce surface water and groundwater withdrawals of thermoelectric power plants throughout the country. Our team’s exploratory work has documented what appears to be a significant and practical potential for augmenting power plant cooling water resources for makeup supply at many, but not all, thermoelectric power plant sites. The intent is to help alleviate stress on existing surface water and groundwater resources through harvesting, storing, polishing and beneficially re-using critical water resources. Through literature review, development of conceptual created wetland plans, and STELLA-based modeling, the AES/SES team has developed heat and water balances for conventional thermoelectric power plants to evaluate wetland size requirements, water use, and comparative cooling technology costs. The ecological literature on organism tolerances to heated waters was used to understand the range of ecological outcomes achievable in created wetlands. This study suggests that wetlands and water harvesting can provide a practical and cost-effective strategy to augment cooling waters for thermoelectric power plants in many geographic settings of the United States, particularly east of the 100th meridian, and in coastal and riverine locations. The study concluded that constructed wetlands can have significant positive

Tokamak experimental power reactor

International Nuclear Information System (INIS)

Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.

1978-01-01

A tokamak experimental power reactor has been designed that is capable of producing net electric power over a wide range of possible operating conditions. A net production of 81 MW of electricity is expected from the design reference conditions that assume a value of 0.07 for beta-toroidal, a maximum toroidal magnetic field of 9 T and a thermal conversion efficiency of 30%. Impurity control is achieved through the use of a low-Z first wall coating. This approach allows a burn time of 60 seconds without the incorporation of a divertor. The system is cooled by a dual pressurized water/steam system that could potentially provide thermal efficiencies as high as 39%. The first surface facing the plasma is a low-Z coated water cooled panel that is attached to a 20 cm thick blanket module. The vacuum boundary is removed a total of 22 cm from the plasma, thereby minimizing the amount of radiation damage in this vital component. Consideration is given in the design to the possible use of the EPR as a materials test reactor. It is estimated that the total system could be built for less than 550 million dollars

Passive residual energy utilization system in thermal cycles on water-cooled power reactors

International Nuclear Information System (INIS)

Placco, Guilherme M.; Guimaraes, Lamartine N.F.; Santos, Rubens S. dos

2013-01-01

This work presents a concept of a residual energy utilization in nuclear plants thermal cycles. After taking notice of the causes of the Fukushima nuclear plant accident, an idea arose to adapt a passive thermal circuit as part of the ECCS (Emergency Core Cooling System). One of the research topics of IEAv (Institute for Advanced Studies), as part of the heat conversion of a space nuclear power system is a passive multi fluid turbine. One of the main characteristics of this device is its passive capability of staying inert and be brought to power at moments notice. During the first experiments and testing of this passive device, it became clear that any small amount of gas flow would generate power. Given that in the first stages of the Fukushima accident and even during the whole event there was plenty availability of steam flow that would be the proper condition to make the proposed system to work. This system starts in case of failure of the ECCS, including loss of site power, loss of diesel generators and loss of the battery power. This system does not requires electricity to run and will work with bleed steam. It will generate enough power to supply the plant safety system avoiding overheating of the reactor core produced by the decay heat. This passive system uses a modified Tesla type turbine. With the tests conducted until now, it is possible to ensure that the operation of this new turbine in a thermal cycle is very satisfactory and it performs as expected. (author)

Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, 3rd Quarterly Report

Energy Technology Data Exchange (ETDEWEB)

Mac Donald, Philip Elsworth

2002-06-01

The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed.

Progress in design, research and development and testing of safety systems for advanced water cooled reactors. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1996-04-01

The meeting covered the following topics: Developments in design of safety-related heat removal components and systems for advanced water cooled reactors; status of test programmes on heat removal components and systems of new designs; range of validity and extrapolation of test results for the qualification of design/licensing computer models and codes for advanced water cooled reactors; future needs and trends in testing of safety systems for advanced water cooled reactors. Tests of heat removal safety systems have been conducted by various groups supporting the design, testing and certification of advanced water cooled reactors. The Technical Committee concluded that the reported test results generally confirm the predicted performance features of the advanced designs. Refs, figs, tabs

Progress in design, research and development and testing of safety systems for advanced water cooled reactors. Proceedings of a technical committee meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-04-01

The meeting covered the following topics: Developments in design of safety-related heat removal components and systems for advanced water cooled reactors; status of test programmes on heat removal components and systems of new designs; range of validity and extrapolation of test results for the qualification of design/licensing computer models and codes for advanced water cooled reactors; future needs and trends in testing of safety systems for advanced water cooled reactors. Tests of heat removal safety systems have been conducted by various groups supporting the design, testing and certification of advanced water cooled reactors. The Technical Committee concluded that the reported test results generally confirm the predicted performance features of the advanced designs. Refs, figs, tabs.

TRACG-CFD analysis of ESBWR reactor water cleanup shutdown cooling system mixing coefficient

International Nuclear Information System (INIS)

Gallardo, J.; Marquino, W.; Mistreanu, A.; Yang, J.

2015-09-01

The ESBWR is a 1520 nominal [M We] Generation III+ natural circulation boiling water reactor designed to high levels of safety utilizing features that have been successfully used before in operating BWRs, as well as standard features common to A BWR. In September of 2014, the US NRC has certified the ESBWR design for use in the USA. The RWCU/Sdc is an auxiliary system for the ESBWR nuclear island. Basic functions it performs include purifying the reactor coolant during normal operation and shutdown and providing shutdown cooling and cooldown to cold shutdown conditions. The performance of the RWCU system during shutdown cooling is directly related to the temperature of the water removed through the outlets, which is coupled with the vessel and F W temperatures through a thermal mixing coefficient. The complex three-dimensional (3-D) geometry of the BWR downcomer and lower plenum has a great impact on the flow mixing. Only a fine mesh technique like CFD can predict the 3-D temperature distribution in the RPV during shutdown and provide the RWCU/Sdc system inlet temperature. Plant shutdown is an unsteady event by nature and was modeled as a succession of CFD steady-state simulations. It is required to establish the mixing coefficient (which is a function of the heat balance and the core flow) during the operation of the RWCU system in the multiple shutdown cooling modes, and therefore a range of core flows needs to be estimated using quasi steady states obtained with TRACG. The lower end of that range is obtained from a system with minimal power decay heat and core flow; while the higher end corresponds to the power at the beginning of RWCU/Sdc operation when the cooldown is transferred to the RWCU/Sdc after the initial depressurization via the turbine bypass valves. Because the ESBWR RWCU/Sdc return and suction designs provide good mixing, the uniform mixing energy balance was found to be an adequate alternative for deriving the mixing coefficient. The CFD mass flow

TRACG-CFD analysis of ESBWR reactor water cleanup shutdown cooling system mixing coefficient

Energy Technology Data Exchange (ETDEWEB)

Gallardo, J. [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Marquino, W.; Mistreanu, A.; Yang, J., E-mail: euqrop@hotmail.com [General Electric Hitachi Nuclear Energy, Wilmington, 28401 North Carolina (United States)

2015-09-15

The ESBWR is a 1520 nominal [M We] Generation III+ natural circulation boiling water reactor designed to high levels of safety utilizing features that have been successfully used before in operating BWRs, as well as standard features common to A BWR. In September of 2014, the US NRC has certified the ESBWR design for use in the USA. The RWCU/Sdc is an auxiliary system for the ESBWR nuclear island. Basic functions it performs include purifying the reactor coolant during normal operation and shutdown and providing shutdown cooling and cooldown to cold shutdown conditions. The performance of the RWCU system during shutdown cooling is directly related to the temperature of the water removed through the outlets, which is coupled with the vessel and F W temperatures through a thermal mixing coefficient. The complex three-dimensional (3-D) geometry of the BWR downcomer and lower plenum has a great impact on the flow mixing. Only a fine mesh technique like CFD can predict the 3-D temperature distribution in the RPV during shutdown and provide the RWCU/Sdc system inlet temperature. Plant shutdown is an unsteady event by nature and was modeled as a succession of CFD steady-state simulations. It is required to establish the mixing coefficient (which is a function of the heat balance and the core flow) during the operation of the RWCU system in the multiple shutdown cooling modes, and therefore a range of core flows needs to be estimated using quasi steady states obtained with TRACG. The lower end of that range is obtained from a system with minimal power decay heat and core flow; while the higher end corresponds to the power at the beginning of RWCU/Sdc operation when the cooldown is transferred to the RWCU/Sdc after the initial depressurization via the turbine bypass valves. Because the ESBWR RWCU/Sdc return and suction designs provide good mixing, the uniform mixing energy balance was found to be an adequate alternative for deriving the mixing coefficient. The CFD mass flow

77 FR 73056 - Initial Test Programs for Water-Cooled Nuclear Power Plants

Science.gov (United States)

2012-12-07

... Plants AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide; request for comment... (DG), DG-1259, ``Initial Test Programs for Water-Cooled Nuclear Power Plants.'' This guide describes... (ITPs) for light water cooled nuclear power plants. DATES: Submit comments by January 31, 2013. Comments...

Chooz A: a model for the dismantling of water-cooled reactors

International Nuclear Information System (INIS)

Anon.

2017-01-01

The specificity of Chooz-A, the first French pressurized water reactor (PWR), is that the reactor and its major components (pumps, exchangers and cooling circuits) are installed in 2 caves dug out in a hill slope. Chooz-A was operating from 1967 to 1991, in 1993 the fuel was removed and in 2007 EDF received the authorization to dismantle the reactor. In 2012, EDF completed the dismantling of the cave containing the elements of the cooling circuit, a cornerstone was the removing of the four 14 m high steam generators. The dismantling of the pressure vessel began in march 2017, it is the same tools and the same processes that were used for the dismantling of the pressure vessel of the Zorita plant (Spain) in 2016. The end of the Chooz-A dismantling is expected in 2022. The feedback experience will help to standardize practices for the French fleet of PWRs. (A.C.)

Emergency cooling method and system for gas-cooled nuclear reactors

International Nuclear Information System (INIS)

Kumpf, H.

1982-01-01

For emergency cooling of gas-cooled fast breeder reactors (GSB), which have a core consisting of a fission zone and a breeding zone, water is sprayed out of nozzles on to the core from above in the case of an incident. The water which is not treated with boron is taken out of a reservoir in the form of a storage tank in such a maximum quantity that the cooling water gathering in the space below the core rises at most up to the lower edge of the fission zone. (orig./GL) [de

Cooling device for reactor suppression pool

International Nuclear Information System (INIS)

Togasaki, Susumu; Kato, Kiyoshi.

1994-01-01

In a cooling device of a reactor suppression pool, when a temperature of pool water is abnormally increased and a heat absorbing portion is heated by, for example, occurrence of an accident, coolants are sent to the outside of the reactor container to actuates a thermally operating portion by the heat energy of coolants and drive heat exchanging fluids of a secondary cooling system. If the heat exchanging fluids are sent to a cooling portion, the coolants are cooled and returned to the heat absorbing portion of the suppression pool water. If the heat absorbing portion is heat pipes, the coolants are evaporated by heat absorbed from the suppression pool water, steams are sent to the thermally operating portion, then coolants are liquefied and caused to return to the heat absorbing portion. If the thermal operation portion is a gas turbine, the gas turbine is operated by the coolants, and it is converted to a rotational force to drive heat exchanging fluids by pumps. By constituting the cooling portion with a condensator, the coolants are condensed and liquefied and returned to the heat absorbing portion of the suppression pool water. (N.H.)

The atmospheric cooling of nuclear power stations

International Nuclear Information System (INIS)

Leuenberger, J.M.; Mayor, J.C.; Gassmann, F.; Lieber, K.

1978-08-01

Four different types of nuclear reactor are considered: light water reactors, high temperature reactors with steam circulation and with direct gas turbine circulation, and fast breeder reactors. Wet and dry cooling towers are described and experimental studies carried out using cooling tower models are presented. (G.T.H.)

Emergency core cooling system in BWR type reactors

International Nuclear Information System (INIS)

Takizawa, Yoji

1981-01-01

Purpose: To rapidly recover the water level in the reactor upon occurrence of slight leakages in the reactor coolant pressure boundary, by promoting the depressurization in the reactor to thereby rapidly increase the high pressure core spray flow rate. Constitution: Upon occurrence of reactor water level reduction, a reactor isolation cooling system and a high pressure core spray system are actuated to start the injection of coolants into a reactor pressure vessel. In this case, if the isolation cooling system is failed to decrease the flow rate in a return pipeway, flow rate indicators show a lower value as compared with a predetermined value. The control device detects it and further confirms the rotation of a high pressure spray pump to open a valve. By the above operation, coolants pumped by the high pressure spray pump is flown by way of a communication pipeway to the return pipeway and sprayed from the top of the pressure vessel. This allows the vapors on the water surface in the pressure vessel to be cooled rapidly and increases the depressurization effects. (Horiuchi, T.)

Renewal of reactor cooling system of JMTR. Reactor building site

International Nuclear Information System (INIS)

Onoue, Ryuji; Kawamata, Takanori; Otsuka, Kaoru; Sekine, Katsunori; Koike, Sumio; Gorai, Shigeru; Nishiyama, Yutaka; Fukasaku, Akitomi

2012-03-01

The Japan Materials Testing Reactor (JMTR) is a light water moderated and cooled tank-type reactor, and its thermal power is 50 MW. The JMTR is categorized as high flux testing reactors in the world. The JMTR has been utilized for irradiation experiments of nuclear fuels and materials, as well as for radioisotope productions since the first criticality in March 1968 until August 2006. JAEA is decided to refurbish the JMTR as an important fundamental infrastructure to promote the nuclear research and development. And The JMTR refurbishment work is carried out for 4 years from 2007. Before refurbishment work, from August 2006 to March 2007, all concerned renewal facilities were selected from evaluation on their damage and wear in terms of aging. Facilities which replacement parts are no longer manufactured or not likely to be manufactured continuously in near future, are selected as renewal ones. Replace priority was decided with special attention to safety concerns. A monitoring of aging condition by the regular maintenance activity is an important factor in selection of continuous using after the restart. In this report, renewal of the cooling system within refurbishment facilities in the JMTR is summarized. (author)

Study on external reactor vessel cooling capacity for advanced large size PWR

International Nuclear Information System (INIS)

Jin Di; Liu Xiaojing; Cheng Xu; Li Fei

2014-01-01

External reactor vessel cooling (ERVC) is widely adopted as a part of in- vessel retention (IVR) in severe accident management strategies. In this paper, some flow parameters and boundary conditions, eg., inlet and outlet area, water inlet temperature, heating power of the lower head, the annular gap size at the position of the lower head and flooding water level, were considered to qualitatively study the effect of them on natural circulation capacity of the external reactor vessel cooling for an advanced large size PWR by using RELAP5 code. And the calculation results provide some basis of analysis for the structure design and the following transient response behavior of the system. (authors)

Safety of next generation power reactors

International Nuclear Information System (INIS)

Anon.

1988-01-01

This book is organized under the following headings: Future needs of utilities regulators, government, and other energy users, PRA and reliability, LMR concepts, LWR design, Advanced reactor technology, What the industry can deliver: advanced LWRs, High temperature gas-cooled reactors, LMR whole-core experiments, Advanced LWR concepts, LWR technology, Forum: public perceptions, What the industry can deliver: LMRs and HTGRs, Criteria and licensing, LMR modeling, Light water reactor thermal-hydraulics, LMR technology, Working together to revitalize nuclear power, Appendix A, luncheon address, Appendix B, banquet address

The 'practical elimination' approach of accident situations for water-cooled nuclear power reactors 2017

International Nuclear Information System (INIS)

2017-01-01

The implementation of the defence in depth principle and current regulations have lead applicants to define provisions to prevent accidents, including severe accidents, and to limit their consequences should they occur. However, while defining the design orientations for a new water-cooled power reactor, applicants shall use the 'practical elimination' approach for severe accident situations (in the reactor core or the spent fuel pool) potentially leading to large early radiological releases, where it appears impossible to define realistic and demonstrable provisions to limit their consequences according to current knowledge and the techniques available at the time. The use of this approach should be discussed between the applicant and the safety authority at the design orientations stage; the authority will specify on a case-by-case basis the conditions for its approval. In order to 'practically eliminate' a situation, the designer shall first examine the possibility for making it physically impossible. Where physical impossibility cannot be achieved, provisions shall be implemented to justify with a high degree of confidence that the situation is extremely unlikely. Situations likely to be 'practically eliminated' are diverse (massive and rapid reactivity insertion accidents, explosions, containment bypasses, etc.); the justification of 'practical elimination' can only be assessed on a case-by-case basis, using deterministic considerations complemented by a probabilistic analysis. The assessment relies on the reactor physical characteristics as well as on the robustness and reliability of the lines of defence implemented to prevent the situation to be 'practically eliminated'. The implemented provisions shall be subject to strong design, manufacturing and operation requirements; considerations related to human factors and hazards shall also be taken into account. This document is an orientation text which defines

77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

Science.gov (United States)

2012-06-15

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling... for public comment draft regulatory guide (DG), DG-1277, ``Initial Test Program of Emergency Core..., entitled, ``Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors,'' is...

Conceptual design tool development for a Pb-Bi cooled reactor

International Nuclear Information System (INIS)

Lee, K. G.; Chang, S. H.; No, H. C.; Chunm, M. H.

2000-01-01

Conceptual design is generally ill-structured and mysterious problem solving. This leads the experienced experts to be still responsible for the most of synthesis and analysis task, which are not amenable to logical formulations in design problems. Especially because a novel reactor such as a Pb-Bi cooled reactor is going on a conceptual design stage, it will be very meaningful to develop the conceptual design tool. This tool consists of system design module with artificial intelligence, scaling module, and validation module. System design decides the optimal structure and the layout of a Pb-Bi cooled reactor, using design synthesis part and design analysis part. The designed system is scaled to be optimal with desired power level, and then the design basis accidents (Dbase) are analyzed in validation module. Design synthesis part contains the specific data for reactor components and the general data for a Pb-Bi cooled reactor. Design analysis part contains several design constraints for formulation and solution of a design problem. In addition, designer's intention may be externalized through emphasis on design requirements. For the purpose of demonstration, the conceptual design tool is applied to a Pb-Bi cooled reactor with 125 M Wth of power level. The Pb-Bi cooled reactor is a novel reactor concept in which the fission-generated heat is transferred from the primary coolant to the secondary coolant through a reactor vessel wall of a novel design. The Pb-Bi cooled reactor is to deliver 125 M Wth per module for 15 effective full power years without any on-site fuel handling. The conceptual design tool investigated the feasibility of a Pb-Bi cooled reactor. Application of the conceptual design tool will be, in detail, presented in the full paper. (author)

Hydrogen production system based on high temperature gas cooled reactor energy using the sulfur-iodine (SI) thermochemical water splitting cycle

International Nuclear Information System (INIS)

Garcia, L.; Gonzalez, D.

2011-01-01

Hydrogen production from water using nuclear energy offers one of the most attractive zero-emission energy strategies and the only one that is practical on a substantial scale. Recently, strong interest is seen in hydrogen production using heat of a high-temperature gas-cooled reactor. The high-temperature characteristics of the modular helium reactor (MHR) make it a strong candidate for producing hydrogen using thermochemical or high-temperature electrolysis (HTE) processes. Eventually it could be also employ a high-temperature gas-cooled reactor (HTGR), which is particularly attractive because it has unique capability, among potential future generation nuclear power options, to produce high-temperature heat ideally suited for nuclear-heated hydrogen production. Using heat from nuclear reactors to drive a sulfur-iodine (SI) thermochemical hydrogen production process has been interest of many laboratories in the world. One of the promising approaches to produce large quantity of hydrogen in an efficient way using the nuclear energy is the sulfur-iodine (SI) thermochemical water splitting cycle. Among the thermochemical cycles, the sulfur iodine process remains a very promising solution in matter of efficiency and cost. This work provides a pre-conceptual design description of a SI-Based H2-Nuclear Reactor plant. Software based on chemical process simulation (CPS) was used to simulate the thermochemical water splitting cycle Sulfur-Iodine for hydrogen production. (Author)

Startup of a high-temperature reactor cooled and moderated by supercritical-pressure light water

International Nuclear Information System (INIS)

Yi, Tin Tin; Ishiwatari, Yuki; Koshizuka, Seiichi; Oka, Yoshiaki

2003-01-01

The startup schemes of high-temperature reactors cooled and moderated by supercritical pressure light water (SCLWR-H) with square lattice and descending flow type water rods are studied by thermal-hydraulic analysis. In this study, two kinds of startup systems are investigated. In the constant pressure startup system, the reactor starts at a supercritical pressure. A flash tank and pressure reducing valves are necessary. The flash tank is designed so that the moisture content in the steam is less than 0.1%. In sliding pressure startup system, the reactor starts at a subcritical pressure. A steam-water separator and a drain tank are required for two-phase flow at startup. The separator is designed by referring to the water separator used in supercritical fossil-fired power plants. The maximum cladding surface temperature during the power-raising phase of startup is restricted not to exceed the rated value of 620degC. The minimum feedwater flow rate is 25% for constant pressure startup and 35% for sliding pressure startup system. It is found that both constant pressure startup system and sliding pressure startup system are feasible in SCLWR-H from the thermal hydraulic point of view. The core outlet temperature as high as 500degC can be achieved in the present design of SCLWR-H. Since the feedwater flow rate of SCLWR-H (1190 kg/s) is lower than that of the previous SCR designs the weight of the component required for startup is reduced. The sliding pressure startup system is better than constant pressure startup system in order to reduce the required component weight (and hence material expenditure) and to simplify the startup plant system. (author)

French activities on gas cooled reactors

International Nuclear Information System (INIS)

Bastien, D.

1996-01-01

The gas cooled reactor programme in France originally consisted of eight Natural Uranium Graphite Gas Cooled Reactors (UNGG). These eight units, which are now permanently shutdown, represented a combined net electrical power of 2,375 MW and a total operational history of 163 years. Studies related to these reactors concern monitoring and dismantling of decommissioned facilities, including the development of methods for dismantling. France has been monitoring the development of HTRs throughout the world since 1979, when it halted its own HTR R and D programme. France actively participates in three CRPs set up by the IAEA. (author). 1 tab

Data on loss of off-site electric power simulation tests of the high temperature engineering test reactor

International Nuclear Information System (INIS)

Takeda, Takeshi; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

2002-07-01

The high temperature engineering test reactor (HTTR), the first high temperature gas-cooled reactor (HTGR) in Japan, achieved the first full power of 30 MW on December 7 in 2001. In the rise-to-power test of the HTTR, simulation tests on loss of off-site electric power from 15 and 30 MW operations were carried out by manual shutdown of off-site electric power. Because helium circulators and water pumps coasted down immediately after the loss of off-site electric power, flow rates of helium and water decreased to the scram points. To shut down the reactor safely, the subcriticality should be kept by the insertion of control rods and the auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components. About 50 s later from the loss of off-site electric power, the auxiliary cooling system started up by supplying electricity from emergency power feeders. Temperature of hot plenum block among core graphite structures decreased continuously after the startup of the auxiliary cooling system. This report describes sequences of dynamic components and transient behaviors of the reactor and its cooling system during the simulation tests from 15 and 30 MW operations. (author)

A dual pressurized water reactor producing 2000 MWe

International Nuclear Information System (INIS)

Kang, K. M.; Suh, K. Y.

2010-01-01

The Dual Unit Optimizer 2000 MWe (DUO2000) is proposed as a new design concept for large nuclear power plant. DUO is being designed to meet economic and safety challenges facing the 21. century green and sustainable energy industry. DUO2000 has two nuclear steam supply systems (NSSSs) of the Unit Nuclear Optimizer (UNO) pressurized water reactor (PWR) in a single containment so as to double the capacity of the plant. UNO is anchored to the Optimized Power Reactor 1000 MWe (OPR1000). The concept of DUO can be extended to any number of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactor (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. In particular, since it is required that the Small and Medium sized Reactors (SMRs) be built as units, the concept of DUO2000 will apply to SMRs as well. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to end, but also pave ways to most promising large power capacity dispensing with huge redesigning cost for Generation III+ nuclear systems. Also, the strengths of DUO2000 include reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS. Two prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The Coolant Unit Branching Apparatus (CUBA) is proposed

Effects of Nuclear Energy on Sustainable Development and Energy Security: Sodium-Cooled Fast Reactor Case

Directory of Open Access Journals (Sweden)

Sungjoo Lee

2016-09-01

Full Text Available We propose a stepwise method of selecting appropriate indicators to measure effects of a specific nuclear energy option on sustainable development and energy security, and also to compare an energy option with another. Focusing on the sodium-cooled fast reactor, one of the highlighted Generation IV reactors, we measure and compare its effects with the standard pressurized water reactor-based nuclear power, and then with coal power. Collecting 36 indicators, five experts select seven key indicators to meet data availability, nuclear energy relevancy, comparability among energy options, and fit with Korean energy policy objectives. The results show that sodium-cooled fast reactors is a better alternative than existing nuclear power as well as coal electricity generation across social, economic and environmental dimensions. Our method makes comparison between energy alternatives easier, thereby clarifying consequences of different energy policy decisions.

European supercritical water cooled reactor (HPLWR Phase 2 project)

International Nuclear Information System (INIS)

Schulenberg, Thomas; Starflinger, Joerg; Marsault, Philippe; Bittermann, Dietmar; Maraczy, Czaba; Laurien, Eckart; Lycklama, Jan Aiso; Anglart, Henryk; Andreani, Michele; Ruzickova, Mariana; Heikinheimo, Liisa

2010-01-01

The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 deg C maximum core outlet temperature. It is designed and analyzed by a European consortium of 13 partners from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small, housed fuel assemblies with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The innovative core design with upward and downward flow through its assemblies has been studied with neutronic, thermal-hydraulic and stress analyses and has been reviewed carefully in a mid-term assessment. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. An overview of results achieved up to now, given in this paper, is illustrating the latest scientific and technological advances. (author)

Passive safety systems and natural circulation in water cooled nuclear power plants

International Nuclear Information System (INIS)

2009-11-01

Nuclear power produces 15% of the world's electricity. Many countries are planning to either introduce nuclear energy or expand their nuclear generating capacity. Design organizations are incorporating both proven means and new approaches for reducing the capital costs of their advanced designs. In the future most new nuclear plants will be of evolutionary design, often pursuing economies of scale. In the longer term, innovative designs could help to promote a new era of nuclear power. Since the mid-1980s it has been recognized that the application of passive safety systems (i.e. those whose operation takes advantage of natural forces such as convection and gravity), can contribute to simplification and potentially improve economics of new nuclear power plant designs. The IAEA Conference on The Safety of Nuclear Power: Strategy for the Future, which was convened in 1991, noted that for new plants 'the use of passive safety features is a desirable method of achieving simplification and increasing the reliability of the performance of essential safety functions, and should be used wherever appropriate'. Some new designs also utilize natural circulation as a means to remove core power during normal operation. The use of passive systems can eliminate the costs associated with the installation, maintenance, and operation of active systems that require multiple pumps with independent and redundant electric power supplies. However, considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to ensure that the systems perform their intended functions. To support the development of advanced water cooled reactor designs with passive systems, investigations of natural circulation are conducted in several IAEA Member States with advanced reactor development programmes. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, the IAEA

Thermohydraulics of emergency core cooling in light water reactors

International Nuclear Information System (INIS)

1989-10-01

This report, by a group of experts of the OECD-NEA Committee on the Safety of Nuclear Installations, reviews the current state-of-knowledge in the field of emergency core cooling (ECC) for design-basis, loss-of-coolant accidents (LOCA) and core uncover transients in pressurized- and boiling-water reactors. An overview of the LOCA scenarios and ECC phenomenology is provided for each type of reactor, together with a brief description of their ECC systems. Separate-effects and integral-test facilities, which contribute to understanding and assessing the phenomenology, are reviewed together with similarity and scaling compromises. All relevant LOCA phenomena are then brought together in the form of tables. Each phenomenon is weighted in terms of its importance to the course of a LOCA, and appraised for the adequacy of its data base and analytical modelling. This qualitative procedure focusses attention on the modelling requirements of dominant LOCA phenomena and the current capabilities of the two-fluid models in two-phase flows. This leads into the key issue with ECC: quantitative code assessment and the application of system codes to predict with a well defined uncertainty the behaviour of a nuclear power plant. This issue, the methodologies being developed for code assessment and the question of how good is good enough are discussed in detail. Some general conclusions and recommendations for future research activities are provided

Three core concepts for producing uranium-233 in commercial pressurized light water reactors for possible use in water-cooled breeder reactors

International Nuclear Information System (INIS)

Conley, G.H.; Cowell, G.K.; Detrick, C.A.; Kusenko, J.; Johnson, E.G.; Dunyak, J.; Flanery, B.K.; Shinko, M.S.; Giffen, R.H.; Rampolla, D.S.

1979-12-01

Selected prebreeder core concepts are described which could be backfit into a reference light water reactor similar to current commercial reactors, and produce uranium-233 for use in water-cooled breeder reactors. The prebreeder concepts were selected on the basis of minimizing fuel system development and reactor changes required to permit a backfit. The fuel assemblies for the prebreeder core concepts discussed would occupy the same space envelope as those in the reference core but contain a 19 by 19 array of fuel rods instead of the reference 17 by 17 array. An instrument well and 28 guide tubes for control rods have been allocated to each prebreeder fuel assembly in a pattern similar to that for the reference fuel assemblies. Backfit of these prebreeder concepts into the reference reactor would require changes only to the upper core support structure while providing flexibility for alternatives in the type of fuel used

Modelling of Transport of Radioactive Substances in the Primary Circuit of Water Cooled Reactors

International Nuclear Information System (INIS)

2012-03-01

Since the beginning of the development of water cooled nuclear power reactors, it has been known that the materials in contact with the water release some of their corrosion products into the water. As a consequence, some of the corrosion products are neutron-activated while in the reactor core and then create a gamma radiation field when deposited outside the core. These radiation fields are hazardous to the inspection, maintenance and operating staff in the power plant and therefore must be minimized. Many methods have been developed to control these radiation fields, such as the proper selection of materials and surface finishing technologies at the design stage, operating and shutdown water chemistry optimization, and the application of different decontamination methods. The need to understand the causes of this radioactivity transport has resulted in many mathematical models to describe the transport, irradiation and deposition of the radioactive corrosion products out of the core. Early models were empirical descriptions of the transport, irradiation and deposition steps, and these models allowed analytical solution of the resulting differential equations. As the mechanisms responsible for radioactivity transport gradually became better understood, more precise models of the mechanisms were made. Computer codes to solve the equations describing these models are necessary. Accurate codes are invaluable design tools for carrying out cost-benefit analysis during materials selection, for estimating shielding thicknesses and for evaluating water chemistry specifications, for example. Such codes are also useful in operating plants to predict radiation fields at specific locations where shielding may be required during a maintenance shutdown, for example, when control of radiation dose to staff is essential. To complement the previous work of the International Atomic Energy Agency (IAEA) to improve the mechanistic understanding of radioactivity transport, a

Device for the condensation of pressurized steam and its application to the cooling of a nuclear reactor after an incident

International Nuclear Information System (INIS)

Dagard, P.; Couturier, M.

1989-01-01

This document describes an invention which relates to a device for condensation of pressurized water which is at a pressure considerably above atmospheric pressure, such as the steam produced by the steam generator of a pressurized-water nuclear reactor during the cooling of the reactor after an incident. The purpose of the invention is therefore to propose a device for the condensation of steam which is under a pressure which is considerably higher than atmospheric pressure by cooling this circulating steam as a result of contact with a heat-exchange wall which is cooled by water; such a device should be easy to install in a nuclear power plant to ensure passive cooling of the reactor, it should have a very good efficiency because of efficient heat exchangers, and it should require only a limited amount of cooling water in the equipment itself

BRENDA: a dynamic simulator for a sodium-cooled fast reactor power plant

International Nuclear Information System (INIS)

Hetrick, D.L.; Sowers, G.W.

1978-06-01

This report is a users' manual for one version of BRENDA (Breeder Reactor Nuclear Dynamic Analysis), which is a digital program for simulating the dynamic behavior of a sodium-cooled fast reactor power plant. This version, which contains 57 differential equations, represents a simplified model of the Clinch River Breeder Reactor Project (CRBRP). BRENDA is an input deck for DARE P (Differential Analyzer Replacement, Portable), which is a continuous-system simulation language developed at the University of Arizona. This report contains brief descriptions of DARE P and BRENDA, instructions for using BRENDA in conjunction with DARE P, and some sample output. A list of variable names and a listing for BRENDA are included as appendices

RBMK power unit operational performance investigation while false coming into action of the emergency reactor cooling system

International Nuclear Information System (INIS)

Emel'yanov, I.Ya.; Aleksakov, A.N.; Vasilevskij, V.P.; Labazov, V.N.; Nikolaev, E.V.; Podlazov, L.N.; Rogov, V.D.; Shevchenko, V.V.

1984-01-01

Regimes of RBMK reactor operation during false coming into action of the emergency reactor cooling system (ERCS), which might occur in the case of faults in the automation systems or erroneous actions of operator have been investigated. At that, thepe exists a probability of water supply from ERCS to one half of the reactor, which results in a sharp change of boiling regime, and due to void reactivity effect it causes the neutron field disturbance. Change in flow rate and enthalpy of coolant, as well as changes of neutron flux in the left and right halves of the reactor at ERCS response and during operation of the whole system of automatic control of power and system of local automatic control of power - local emergency protection - have been studied. The investigations have been carried out for different values of vapour effect of void reactivity effect and for time ranges from 0 to 40 s. The calculations are made using a model, describing spatial dynamics of reactor in two-dimensional approximation with 54 nodes. The model describes neutron-physics and thermohydraulic processes and it is realized using the BEhSM-6 computer. It is pointed out that one system of automatic control of power or local emergency protection (i.e. without the shut-off system), is insufficient for the compensation of disturbances appearing as a result of false ERCS coming into operation

Natural circulation in water cooled nuclear power plants: Phenomena, models, and methodology for system reliability assessments

International Nuclear Information System (INIS)

2005-11-01

In recent years it has been recognized that the application of passive safety systems (i.e. those whose operation takes advantage of natural forces such as convection and gravity), can contribute to simplification and potentially to improved economics of new nuclear power plant designs. Further, the IAEA Conference on The Safety of Nuclear Power: Strategy for the Future which was convened in 1991 noted that for new plants 'the use of passive safety features is a desirable method of achieving simplification and increasing the reliability of the performance of essential safety functions, and should be used wherever appropriate'. Considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to assure that the systems perform their intended functions. To support the development of advanced water cooled reactor designs with passive systems, investigations of natural circulation are an ongoing activity in several IAEA Member States. Some new designs also utilize natural circulation as a means to remove core power during normal operation. In response to the motivating factors discussed above, and to foster international collaboration on the enabling technology of passive systems that utilize natural circulation, an IAEA Coordinated Research Project (CRP) on Natural Circulation Phenomena, Modelling and Reliability of Passive Systems that Utilize Natural Circulation was started in early 2004. Building on the shared expertise within the CRP, this publication presents extensive information on natural circulation phenomena, models, predictive tools and experiments that currently support design and analyses of natural circulation systems and highlights areas where additional research is needed. Therefore, this publication serves both to provide a description of the present state of knowledge on natural circulation in water cooled nuclear power plants and to guide the planning and conduct of the CRP in

Method and plant to remote tritium from the cooling water of a nuclear reactor

International Nuclear Information System (INIS)

O'Brien, C.J.

1976-01-01

A method is proposed for the extraction of tritium from the cooling water of a nuclear reactor, based on the principle of concentrating the tritium by a multi-stage transfer process. The cooling water is brought into contact in each stage with basic, labile, hydrogen-containing material with high pH value, whereby the tritium is transfered into an intermediate solid product and can be separated off. The technical details of the plant are described. Cellulose materials, such as cotton and wood as well as protein-containing material, such as muscle tissue are mentioned as examples of materials with a high affinity to tritium, greater than the affinity of water to tritium. They extract tritium from the cooling water. (HK) [de

Refurbishment of the Primary Cooling System of the Puspati Triga Reactor

Energy Technology Data Exchange (ETDEWEB)

Ramli, S.; Zakaria, M. F.; Masood, Z. [Malaysian Nuclear Agency, Kajang (Malaysia)

2014-08-15

The refurbishment of the 27 year old primary cooling system of the 1 MW PUSPATI TRIGA reactor was completed in April 2010 over an eight month outage. The project was implemented with the dual objective of meeting current user needs as well as a future reactor core power upgrade. Hence the cooling system was partly modernized to cater for a 3 MW{sub th} reactor by installing higher capacity heat exchangers and pumps while maintaining the piping and valve sizes. The old 1 MW tube and shell heat exchanger, which had lost 25% of its heat exchange capacity, was replaced with two 1.5 MW plate type heat exchangers. Several manually operated valves were replaced with motorized units to allow remote operation from the control room. The installed cooling system was flushed with distilled water and then subjected to hydrostatic pressure tests. In the cold run test, the system was operated for an hour for every pump and heat exchanger combination while all operating parameters were checked. In the hot run test, the same was done at four levels of increasing reactor power, and dose measurements were also recorded. The paper gives the design, installation, testing and commissioning details of the project. (author)

Performance of materials in the component cooling water systems of pressurized water reactors

International Nuclear Information System (INIS)

Lee, B.S.

1993-01-01

The component cooling water (CCW) system provides cooling water to several important loads throughout the plant under all operating conditions. An aging assessment CCW systems in pressurized water reactors (PWRs) was conducted as part of Nuclear Plant Aging Research Program (NPAR) instituted by the US Nuclear Regulatory Commission. This paper presents some of the results on the performances of materials in respect of their application in CCW Systems. All the CCW system failures reported to the Nuclear Plant Reliability Data System (NPRDS) from January 1988 to June 1990 were reviewed; it is concluded that three of the main contributors to CCW system failures are valves, pumps, and heat exchangers. This study identified the modes and causes of failure for these components; most of the causes for the aging-related failures could be related to the performance of materials. Also, in this paper the materials used for these components are reviewed, and there aging mechanisms under CCW system conditions are discussed

Water-ingress analysis for the 200 MWe pebble-bed modular high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Zheng Yanhua; Shi Lei; Wang Yan

2010-01-01

Water ingress into the primary circuit is generally recognized as one of the severe accidents with potential hazard to the modular high temperature gas-cooled reactor adopting steam-turbine cycle, which will cause a positive reactivity introduction, as well as the chemical corrosion of graphite fuel elements and reflector structure material. Besides, increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes and flammable water gas. The analysis of such a kind of important and particular accident is significant to verify the inherent safety characteristics of the modular HTR plants. Based on the preliminary design of the 200 MWe high temperature gas-cooled reactor pebble-bed modular (HTR-PM), the design basis accident of a double-ended guillotine break of one heating tube and the beyond design basis accident of a large break of the main steam collection plate have been analyzed by using TINTE code, which is a special transient analysis program for high temperature gas-cooled reactors. Some safety relevant concerns, such as the fuel temperature, the primary loop pressure, the graphite corrosion, the water gas releasing amount, as well as the natural convection influence on the condition of failing to close the blower flaps, have been studied in detail. The calculation results indicate that even under some severe hypothetical postulates, the HTR-PM is able to keep the inherent safeties of the modular high temperature gas-cooled reactor and has a relatively good natural plant response, which will not result in environmental radiation hazard.

Emergency cooling device for reactors

International Nuclear Information System (INIS)

Inoue, Hisamichi; Naito, Masanori; Sato, Chikara; Chino, Koichi.

1975-01-01

Object: To pour high pressure cooling water into a core, when coolant is lost in a boiling water reactor, thereby restraining the rise of fuel cladding. Structure: A control rod guiding pipe, which is moved up and down by a control rod, is mounted on the bottom of a pressure vessel, the control rod guiding pipe being communicated with a high pressure cooling water tank positioned externally of the pressure vessel, and a differential in pressure between the pressure vessel and the aforesaid tank is detected when trouble of coolant loss occurs, and the high pressure cooling water within the tank is poured into the core through the control rod guiding pipe to restrain the rise of fuel cladding. (Kamimura, M.)

Technology of steam generators for gas-cooled reactors. Proceedings of a specialists' meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1988-07-01

The activity of the IAEA in the field of the technology of gas-cooled reactors was formalized by formation of an International Working Group on Gas-Cooled Reactors (IWGCR). The gas cooled reactor program considered by the IWGCR includes carbon-dioxide-cooled thermal reactors, helium cooled thermal high temperature reactors for power generation and for process heat applications and gas-cooled fast breeder reactors. This report covers the papers dealing with operating experience, steam generators for next generation of gas-cooled reactors, material development and corrosion problems, and thermohydraulics.

Technology of steam generators for gas-cooled reactors. Proceedings of a specialists' meeting

International Nuclear Information System (INIS)

1988-01-01

The activity of the IAEA in the field of the technology of gas-cooled reactors was formalized by formation of an International Working Group on Gas-Cooled Reactors (IWGCR). The gas cooled reactor program considered by the IWGCR includes carbon-dioxide-cooled thermal reactors, helium cooled thermal high temperature reactors for power generation and for process heat applications and gas-cooled fast breeder reactors. This report covers the papers dealing with operating experience, steam generators for next generation of gas-cooled reactors, material development and corrosion problems, and thermohydraulics

Candidate Materials Evaluation for Supercritical Water-Cooled Reactor

International Nuclear Information System (INIS)

Allen, T.R.; Was, G.S.

2008-01-01

Final technical report on the corrosion, stress corrosion cracking, and radiation response of candidate materials for the supercritical water-cooled reactor concept. The objective of the proposed research was to investigate degradation of materials in the supercritical water environment (SCW). First, representative alloys from the important classes of candidate materials were studied for their corrosion and stress-corrosion cracking (SCC) resistance in supercritical water. These included ferritic/martensitic (F/M) steels, austenitic stainless steels, and Ni-base alloys. Corrosion and SCC tests were conducted at various temperatures and exposure times, as well as in various water chemistries. Second, emerging plasma surface modification and grain boundary engineering technologies were applied to modify the near surface chemistry, microstructure, and stress-state of the alloys prior to corrosion testing. Third, the effect of irradiation on corrosion and SCC of alloys in the as-received and modified/engineered conditions were examined by irradiating samples using high-energy protons and then exposing them to SCW

Modular high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Shepherd, L.R.

1988-01-01

The high financial risk involved in building large nuclear power reactors has been a major factor in halting investment in new plant and in bringing further technical development to a standstill. Increased public concern about the safety of nuclear plant, particularly after Chernobyl, has contributed to this stagnation. Financial and technical risk could be reduced considerably by going to small modular units, which would make it possible to build up power station capacity in small steps. Such modular plant, based on the helium-cooled high temperature reactor (HTR), offers remarkable advantages in terms of inherent safety characteristics, partly because of the relatively small size of the individual modules but more on account of the enormous thermal capacity and high temperature margins of the graphitic reactor assemblies. Assessments indicate that, in the USA, the cost of power from the modular systems would be less than that from conventional single reactor plant, up to about 600 MW(e), and only marginally greater above that level, a margin that should be offset by the shorter time required in bringing the modular units on line to earn revenue. The modular HTR would be particularly appropriate in the UK, because of the considerable British industrial background in gas-cooled reactors, and could be a suitable replacement for Magnox. The modular reactor would be particularly suited to combined heat and power schemes and would offer great potential for the eventual development of gas turbine power conversion and the production of high-temperature process heat. (author)

Predicted effect of power uprating on the water chemistry of commercial boiling water reactors

International Nuclear Information System (INIS)

Yeh, Tsung-Kuang; Wang, Mei-Ya; Chu, Charles F.; Chang Ching

2009-01-01

The approach of power uprating has been adopted by operators of light water reactors in the past few decades in order to increase the power generation efficiency of nuclear reactors. The power uprate strategy is apparently applicable to the three nuclear reactors in Taiwan as well. When choosing among the three types of power uprating, measurement uncertainty, stretch power uprating, and extended power uprating, a deliberate and thorough evaluation is required before a final decision and an optimal selection can be made. One practical way of increasing the reactor power is to deliberately adjust the fuel loading pattern and the control rod pattern and thus to avoid replacing the primary coolant pump with a new one of larger capacity. The power density of the reactor will increase with increasing power, but the mass flow rate in the primary coolant circuit (PCC) of a light water reactor will slightly increase (usually by less than 5 %) or even remain unchanged. Accordingly, an uprated power would induce higher neutron and gamma photon dose rates in the reactor coolant but have a minor or no effect on the mass flow rate of the primary coolant. The radiolysis product concentrations and the electrochemical corrosion potential (ECP) values differ largely in the PCC of a boiling water reactor (BWR). It is very difficult to measure the water chemistry data directly at various locations of an actual reactor. Thus the impact of power uprating on the water chemistry of a BWR operating under hydrogen water chemistry (HWC) can only be theoretically evaluated through computer modelling. In this study, the DEMACE computer code was modified to investigate the impact of power uprating on the water chemistry under a fixed mass flow rate in the primary coolant circuit of a BWR/6 type plant. Simulations were carried out for hydrogen concentrations in feedwater ranging from 0.0 to 2.0 mg . kg -1 and for power levels ranging from 100 % to 120 %. The responses of water chemistry and ECP

Status of control assembly materials in Indian water reactors

International Nuclear Information System (INIS)

Date, V.G.; Kulkarni, P.G.

2000-01-01

India's present operating water cooled power reactors comprise boiling water reactors of Tarapur Atomic Power Station (TAPS) and pressurized heavy water reactors (PHWRs) at Kota (RAPS), Kalpakkam (MAPS), Narora (NAPS) and Kakrapara (KAPS). Boiling water reactors of TAPS use boron carbide control blades for control of power as well as for shut down (scram). PHWRs use boron steel and cobalt absorber rods for power control and Cd sandwiched shut off rods (primary shut down system) and liquid poison rods (secondary shut down system) for shut down. In TAPS, Gadolinium rods (burnable poison rods) are also incorporated in fuel assembly for flux flattening. Boron carbide control blades and Gadolinium rods for TAPS, cobalt absorber rods and shut down assemblies for PHWRs are fabricated indigenously. Considerable development work was carried out for evolving material specifications, component and assembly drawings, and fabrication processes. Details of various control and shut off assemblies being fabricated currently are highlighted in the paper. (author)

HAZOP-study on heavy water research reactor primary cooling system

International Nuclear Information System (INIS)

Hashemi-Tilehnoee, M.; Pazirandeh, A.; Tashakor, S.

2010-01-01

By knowledge-based Hazard and Operability (HAZOP) technique, equipment malfunction and deficiencies in the primary cooling system of the generic heavy water research reactor are studied. This technique is used to identify the representative accident scenarios. The related Process Flow Drawing (PFD) is prepared as our study database for this plant. Since this facility is in the design stage, applying the results of HAZOP-study to PFD improves the safety of the plant.

Data on test results of vessel cooling system of high temperature engineering test reactor

International Nuclear Information System (INIS)

Saikusa, Akio; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo