Corrosion of research reactor aluminium clad spent fuel in water. Additional information

International Nuclear Information System (INIS)

2009-12-01

A large variety of research reactor spent fuel with different fuel meats, different geometries and different enrichments in 235 U are presently stored underwater in basins located around the world. More than 90% of these fuels are clad in aluminium or aluminium based alloys that are notoriously susceptible to corrosion in water of less than optimum quality. Some fuel is stored in the reactor pools themselves, some in auxiliary pools (or basins) close to the reactor and some stored at away-from-reactor pools. Since the early 1990s, when corrosion induced degradation of the fuel cladding was observed in many of the pools, corrosion of research reactor aluminium clad spent nuclear fuel stored in light water filled basins has become a major concern, and programmes were implemented at the sites to improve fuel storage conditions. The IAEA has since then established a number of programmatic activities to address corrosion of research reactor aluminium clad spent nuclear fuel in water. Of special relevance was the Coordinated Research Project (CRP) on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase I) initiated in 1996, whose results were published in IAEA Technical Reports Series No. 418. At the end of this CRP it was considered necessary that a continuation of the CRP should concentrate on fuel storage basins that had demonstrated significant corrosion problems and would therefore provide additional insight into the fundamentals of localized corrosion of aluminium. As a consequence, the IAEA started a new CRP entitled Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase II), to carry out more comprehensive research in some specific areas of corrosion of aluminium clad spent nuclear fuel in water. In addition to this CRP, one of the activities under IAEA's Technical Cooperation Regional Project for Latin America Management of Spent Fuel from Research Reactors (2001-2006) was corrosion monitoring and surveillance of research

Pressurized water reactor inspection procedures

International Nuclear Information System (INIS)

Heinrich, D.; Mueller, G.; Otte, H.J.; Roth, W.

1998-01-01

Inspections of the reactor pressure vessels of pressurized water reactors (PWR) so far used to be carried out with different central mast manipulators. For technical reasons, parallel inspections of two manipulators alongside work on the refueling cavity, so as to reduce the time spent on the critical path in a revision outage, are not possible. Efforts made to minimize the inspection time required with one manipulator have been successful, but their effects are limited. Major reductions in inspection time can be achieved only if inspections are run with two manipulators in parallel. The decentralized manipulator built by GEC Alsthom Energie and so far emmployed in boiling water reactors in the USA, Spain, Switzerland and Japan allows two systems to be used in parallel, thus reducing the time required for standard inspection of a pressure vessel from some six days to three days. These savings of approximately three days are made possible without any compromises in terms of positioning by rail-bound systems. During inspection, the reactor refueling cavity is available for other revision work without any restrictions. The manipulator can be used equally well for inspecting standard PWR, PWR with a thermal shield, for inspecting the land between in-core instrumentation nozzles, BWR with and without jet pumps (complementary inspection), and for inspecting core support shrouds. (orig.) [de

The potential importance of water pathways for spent fuel transportation accident risk

International Nuclear Information System (INIS)

Ostmeyer, R.M.

1986-01-01

This paper analyzes the potential importance of water pathway contamination for spent fuel transportation accident risk using a ''worst-case'' water contamination scenario. The scenario used for the analysis involves an accident release that occurs near a reservoir. Water pathway doses are compared to doses for accident releases in urban or agricultural areas. The results of the analysis indicate that water pathways are not important for assessing the risk of transporting spent reactor fuel by truck or by rail

Test plan for reactions between spent fuel and J-13 well water under unsaturated conditions

International Nuclear Information System (INIS)

Finn, P.A.; Wronkiewicz, D.J.; Hoh, J.C.; Emery, J.W.; Hafenrichter, L.D.; Bates, J.K.

1993-01-01

The Yucca Mountain Site Characterization Project is evaluating the long-term performance of a high-level nuclear waste form, spent fuel from commercial reactors. Permanent disposal of the spent fuel is possible in a potential repository to be located in the volcanic tuff beds near Yucca Mountain, Nevada. During the post-containment period the spent fuel could be exposed to water condensation since of the cladding is assumed to fail during this time. Spent fuel leach (SFL) tests are designed to simulate and monitor the release of radionuclides from the spent fuel under this condition. This Test Plan addresses the anticipated conditions whereby spent fuel is contacted by small amounts of water that trickle through the spent fuel container. Two complentary test plans are presented, one to examine the reaction of spent fuel and J-13 well water under unsaturated conditions and the second to examine the reaction of unirradiated UO 2 pellets and J-13 well water under unsaturated conditions. The former test plan examines the importance of the water content, the oxygen content as affected by radiolysis, the fuel burnup, fuel surface area, and temperature. The latter test plant examines the effect of the non-presence of Teflon in the test vessel

Reactor-specific spent fuel discharge projections: 1985 to 2020

International Nuclear Information System (INIS)

Heeb, C.M.; Libby, R.A.; Walling, R.C.; Purcell, W.L.

1986-09-01

The creation of four spent-fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No New Orders with Extended Burnup, (2) No New Orders with Constant Burnup, (3) Middle Case with Extended Burnup, and (4) Middle Case with Constant Burnup. Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel

Reactor-specific spent fuel discharge projections, 1987-2020

International Nuclear Information System (INIS)

Walling, R.C.; Heeb, C.M.; Purcell, W.L.

1988-03-01

The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from U.S. commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water eactors (BWR), and one existing high temperature gas reactor (HTGR). The projections are based on individual reactor information supplied by the U.S. reactor owners. The basic information is adjusted to conform to Energy Information Administration (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: No New Orders (assumes increasing burnup), No New Orders with No Increased Burnup, Upper Reference (assumes increasing burnup), Upper Reference with No Increased Burnup, and Lower Reference (assumes increasing burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum at-reactor storage, and for storage requirements assuming maximum at-reactor storage plus intra-utility transshipment of spent fuel. 8 refs., 8 figs., 10 tabs

Criticality safety of spent fuel casks considering water inleakage

International Nuclear Information System (INIS)

Osgood, N.L.; Withee, C.J.; Easton, E.P.

2004-01-01

A fundamental safety design parameter for all fissile material packages is that a single package must be critically safe even if water leaks into the containment system. In addition, criticality safety must be assured for arrays of packages under normal conditions of transport (undamaged packages) and under hypothetical accident conditions (damaged packages). The U.S. Nuclear Regulatory Commission staff has revised the review protocol for demonstrating criticality safety for spent fuel casks. Previous review guidance specified that water inleakage be considered under accident conditions. This practice was based on the fact that the leak tightness of spent fuel casks is typically demonstrated by use of structural analysis and not by physical testing. In addition, since a single package was shown to be safe with water inleakage, it was concluded that this analysis was also applicable to an array of damaged packages, since the heavy shield walls in spent fuel casks neutronically isolate each cask in the array. Inherent in this conclusion is that the fuel assembly geometry does not change significantly, even under drop test conditions. Requests for shipping fuel with burnup exceeding 40 GWd/MTU, including very high burnups exceeding 60 GWD/MTU, caused a reassessment of this assumption. Fuel cladding structural strength and ductility were not clearly predictable for these higher burnups. Therefore the single package analysis for an undamaged package may not be applicable for the damaged package. NRC staff developed a new practice for review of spent fuel casks under accident conditions. The practice presents two methods for approval that would allow an assessment of potential reconfiguration of the fuel assembly under accident conditions, or, alternatively, a demonstration of the water-exclusion boundary through physical testing

Dataset on the cost estimation for spent filter backwash water (SFBW treatment

Directory of Open Access Journals (Sweden)

Afshin Ebrahimi

2017-12-01

Full Text Available The dataset presented in this article are related to the research article entitled “Hybrid coagulation-UF processes for spent filter backwash water treatment: a comparison studies for PAFCl and FeCl3 as a pre-treatment” (Ebrahimi et al., 2017 [1]. This article reports the cost estimation for treating produced spent filter backwash water (SFBW during water treatment in Isfahan- Iran by various methods including primary sedimentation, coagulation & flocculation, second clarification, ultra filtration (UF and recirculation of settled SFBW to water treatment plant (WTP entrance. Coagulation conducted by PAFCl and FeCl3 as pre polymerized and traditional coagulants. Cost estimation showed that contrary to expectations, the recirculation of settled SFBW to WTP entrance is more expensive than other method and it costs about $ 37,814,817.6. Versus the cheapest option related to separate primary sedimentation, coagulation & flocculation in WTP. This option cost about $ 4,757,200 and $ 950,213 when FeCl3 and PAFCl used as coagulant, respectively. Keywords: Spent filter backwash water, Water treatment, Coat estimation, Water reuse

Spent fuel data base: commercial light water reactors. [PWR; BWR

Energy Technology Data Exchange (ETDEWEB)

Hauf, M.J.; Kniazewycz, B.G.

1979-12-01

As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

Reactor-specific spent fuel discharge projections: 1986 to 2020

International Nuclear Information System (INIS)

Heeb, C.M.; Walling, R.C.; Purcell, W.L.

1987-03-01

The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No new orders with extended burnup, (2) No new orders with constant burnup, (3) Upper reference (which assumes extended burnup), (4) Upper reference with constant burnup, and (5) Lower reference (which assumes extended burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel. 6 refs., 8 figs., 8 tabs

Feasibility assessment of burnup credit in the criticality analysis of shipping casks with boiling water reactor spent fuel

International Nuclear Information System (INIS)

Broadhead, B.L.

1991-08-01

Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent fuel casks used for transportation and storage. Recently, analyses have demonstrated the technical feasibility and estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This report summarizes the extension of the previous PWR technical feasibility assessment to boiling water reactor (BWR) fuel. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. Two different aspects of fuel characterization were considered:l first, the generation of burn- up dependent material interaction probabilities; second, the prediction of material inventories over time (depletion). After characterizing the spent fuel at various stages of exposure and decay, three dimensional (3-D) models for an infinite array of assemblies and, in several cases, infinite arrays of assemblies in a typical shipping cask basket were analyzed. Results for assemblies without a basket provide reactivity control requirements as a function of burnup and decay, while results including the basket allow assessment of typical basket configurations to provide sufficient reactivity control for spent BWR fuel. Resulting basket worths and reactivity trends over time are then evaluated to determine whether burnup credit is needed and feasible in BWR applications

Properties of light water reactor spent fuel cladding. Interim report

International Nuclear Information System (INIS)

Farwick, D.G.; Moen, R.A.

1979-08-01

The Commercial Waste and Spent Fuel Packaging Program will provide containment packages for the safe storage or disposal of spent Light Water Reactor (LWR) fuel. Maintaining containment of radionuclides during transportation, handling, processing and storage is essential, so the best understanding of the properties of the materials to be stored is necessary. This report provides data collection, assessment and recommendations for spent LWR fuel cladding materials properties. Major emphasis is placed on mechanical properties of the zircaloys and austenitic stainless steels. Limited information on elastic constants, physical properties, and anticipated corrosion behavior is also provided. Work is in progress to revise these evaluations as the program proceeds

Spent nuclear fuel transportation casks evaluation for water in-leakage

International Nuclear Information System (INIS)

Shah, M.J.; Huang, D.T.; Guttmann, J.; Klymyshyn, N.A.; Koeppel, B.J.; Adkins, H.E.

2004-01-01

The United States Nuclear Regulatory Commission (USNRC) is responsible for licensing commercial spent fuel storage and transportation systems. To ensure that the regulations are risk-informed, and do not place unnecessary regulatory burden on the industry, the USNRC has been examining its regulations that apply to spent fuel transportation casks for maintaining sub-criticality under hypothetical accident conditions. Code of Federal Regulations, Title 10, Part 71[1] (10 CFR 71), section 71.55(b) requires that, for evaluation of sub-criticality for fissile material packages, water moderation should be assumed to occur to the most reactive credible extent consistent with the chemical and physical form of the contents. This requirement is based on a defense-in-depth policy, and accounts for any possibility of water intrusion into the package. This program is designed to quantify the margins of safety of certified transportation casks to water intrusion following hypothetical accident conditions. This paper describes the current status of analytical work being performed to evaluate two USNRC-certified spent fuel transportation casks, HI-STAR 100[2] and TN-68[3]. The analytical work is performed using the ANSYS registered [4] and LS-DYNA trademark [5] finite element analysis (FEA) codes. The models are sufficiently detailed in the areas of bolt closure interfaces and containment boundaries to evaluate the likelihood water in-leakage under free drop hypothetical accident conditions of 10 CFR 71.73

Standard guide for drying behavior of spent nuclear fuel

CERN Document Server

American Society for Testing and Materials. Philadelphia

2008-01-01

1.1 This guide is organized to discuss the three major components of significance in the drying behavior of spent nuclear fuel: evaluating the need for drying, drying spent nuclear fuel, and confirmation of adequate dryness. 1.1.1 The guide addresses drying methods and their limitations in drying spent nuclear fuels that have been in storage at water pools. The guide discusses sources and forms of water that remain in SNF, its container, or both, after the drying process and discusses the importance and potential effects they may have on fuel integrity, and container materials. The effects of residual water are discussed mechanistically as a function of the container thermal and radiological environment to provide guidance on situations that may require extraordinary drying methods, specialized handling, or other treatments. 1.1.2 The basic issue in drying is to determine how dry the SNF must be in order to prevent issues with fuel retrievability, container pressurization, or container corrosion. Adequate d...

Water Quality Analysis Study Pond and Interim Storage for Spent Fuel

International Nuclear Information System (INIS)

Dyah Sulistyani R; Husen Zamroni; Sudiyati

2007-01-01

Purification system of Storage facility of spent fuel which there is in Indonesia is integrated purification system. Reservoir pond of fuel contains approximately 995 m 3 demin water and in pond equipped with some of reservoir racks of spent fuel which must always avoid from factor-factor causing corrosion. In process of this purification system, water impurity which has been activation and also which is not is activation before will filtered and catch by passing of ion exchange so that will reduce conductivity and fuel coolant water activity. Water quality pond and canals links must fulfill specifications, among other: degree of acidity (pH) primary cooling water ranges from 5.5 and 6.5 ; its conductivity 1 - 8 μ S/cm, content analysis CI 0.03 - 0.06 ppm and NO 3 0.1 - 0.2 ppm, radionuclide activity Cs 137 742 Bq/l and Co 60 657 Bq/l and the temperature be kept of less than 40℃ to avoid from corrosion speed. (author)

Effect of water α radiolysis on the spent nuclear fuel UO2 matrix alteration

International Nuclear Information System (INIS)

Lucchini, J.F.

2001-01-01

In the option of long term storage or direct disposal of nuclear spent fuel, it is essential to study the long-term behaviour of the spent fuel matrix (UO 2 ) in water, in presence of ionizing radiations. This work gives some knowledge elements about the impact of aerated water alpha radiolysis on UO 2 alteration. An original experiment method was used in this study. UO 2 /water interfaces were irradiated by an external He 2+ ions beam. The sequential batch dissolution tests on UO 2 samples were performed in aerated deionized water, before, during and after a-irradiation under high fluxes. A corrosion product, identified as hydrated uranium peroxide, was formed on the UO 2 surface. The uranium release was 3 to 4 orders of magnitude higher under irradiation than out of irradiation. The concentrations of the radiolysis products H 2 O 2 and H 3 O + were affected by the uranium oxide surface. They could not only explain the whole uranium release reached during irradiation in water. Leaching experiments on UO X spent fuel samples (with or without the Zircaloy clad) were also performed, in hot cells. The uranium release was relatively small, and H 2 O 2 was not detected in solution. The rates of uranium release in aerated water during one hour were calculated. They were about mg -1 .m -2 .d -1 for spent fuel and for UO 2 , and about g -1 .m -2 .d -1 for UO 2 irradiated by He 2+ ions. The comparison of the results between the two kinds of experiment shows a difference of the behaviour in water between UO 2 irradiated by He 2+ ions and spent fuel. Some hypothesis are given to explain this difference. (author)

Design of make-up water system for Tehran research reactor spent nuclear fuels storage pool

Energy Technology Data Exchange (ETDEWEB)

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

2010-10-15

Spent nuclear fuels storage (SNFS) is an essential auxiliary system in nuclear facility. Following discharge from a nuclear reactor, spent nuclear fuels have to be stored in water pool of SNFS away from reactor to allow for radioactive to decay and removal of generated heat. To prevent corrosion damage of fuels and other equipments, the storage pool is filled with de-ionized water which serves as moderator, coolant and shielding. The de-ionized water will be provided from make-up water system. In this paper, design of a make-up water system for optimal water supply and its chemical properties in SNFS pool is presented. The main concern of design is to provide proper make-up water throughout the storage time. For design of make-up water system, characteristics of activated carbon purifier, anionic, cationic and mixed-bed ion-exchangers have been determined. Inlet water to make-up system provide from Tehran municipal water system. Regulatory Guide 1.13 of the and graver company manual that manufactured the Tehran research reactor (TRR) make-up water system have been used for make-up water system of TRR spent nuclear fuels storage pool design.

Design of make-up water system for Tehran research reactor spent nuclear fuels storage pool

International Nuclear Information System (INIS)

Aghoyeh, Reza Gholizadeh; Khalafi, Hosein

2010-01-01

Spent nuclear fuels storage (SNFS) is an essential auxiliary system in nuclear facility. Following discharge from a nuclear reactor, spent nuclear fuels have to be stored in water pool of SNFS away from reactor to allow for radioactive to decay and removal of generated heat. To prevent corrosion damage of fuels and other equipments, the storage pool is filled with de-ionized water which serves as moderator, coolant and shielding. The de-ionized water will be provided from make-up water system. In this paper, design of a make-up water system for optimal water supply and its chemical properties in SNFS pool is presented. The main concern of design is to provide proper make-up water throughout the storage time. For design of make-up water system, characteristics of activated carbon purifier, anionic, cationic and mixed-bed ion-exchangers have been determined. Inlet water to make-up system provide from Tehran municipal water system. Regulatory Guide 1.13 of the and graver company manual that manufactured the Tehran research reactor (TRR) make-up water system have been used for make-up water system of TRR spent nuclear fuels storage pool design.

Corrosion of research reactor aluminium-clad spent fuel in water-chemical and microbiological influenced

International Nuclear Information System (INIS)

Maksin, T.N.; Dobrijevic, R.P.; Idjakovic, Z.E.; Pesic, M.P.

2002-01-01

Spent fuel resulting from 25 years of operating research reactor RA at the Vinca Institute is presently all stored in the temporary spent fuel storage pool. It has been left in the ambient temperature and humidity for more then fifteen years so intensive corrosion processes were notice. We have spent fuel pools under control, after first research coordination meeting (RCM), of the first CRP, by monitoring of physical and chemical parameters of water in the pools, including temperature, pH-factor, electrical conductivity, mass concentration of corrosion products in the water and mud, mass concentration of relevant ions etc. The rack of standard corrosion coupons, was given at that time, has been in poor quality water for six years. We pick up rack assembly from basin and analysed. The results of this investigation are present in this article. (author)

Corrosion of aluminium alloy test coupons in water of spent fuel storage pool at RA reactor

International Nuclear Information System (INIS)

Pesic, M.; Maksin, T.; Jordanov, G.; Dobrijevic, R.

2004-12-01

Study on corrosion of aluminium cladding, of the TVR-S type of enriched uranium spent fuel elements of the research reactor RA in the storage water pool is examined in the framework nr the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) 'Corrosion of Research Reactor Clad-Clad Spent Fuel in Water' since 2002. Standard racks with aluminium coupons are exposed to water in the spent fuel pools of the research reactor RA. After predetermined exposure times along with periodic monitoring of the water parameters, the coupons are examined according to the strategy and the protocol supplied by the IAEA. Description of the standard corrosion racks, experimental protocols, test procedures, water quality monitoring and compilation of results of visual examination of corrosion effects are present in this article. (author)

Alteration mechanisms of UOX spent fuel under water

International Nuclear Information System (INIS)

Muzeau, B.

2008-06-01

The mechanisms of spent fuel alteration in aqueous media need to be understood on the assumption of a direct disposal of the assemblies in a geological formation or for long duration storage in pool. This work is a contribution to the study of the effects of the alpha and/or beta/gamma radiolysis of water on the oxidation and the dissolution of the UO 2 matrix of UOX spent fuel. The effects of the alpha radiolysis, predominant in geological disposal conditions, were quantified by using samples of UO 2 doped with plutonium. The leaching experiments highlighted two types of control for the matrix alteration according to the alpha activity. The first is based on the radiolytic oxidation of the surface and leads to a continuous release of uranium in solution whereas the second is based on a control by the solubility of uranium. An activity threshold, between 18 MBq.g -1 and 33 MBq.g -1 , was defined in a carbonated water. The value of this threshold is dependent on the experimental conditions and the presence or not of electro-active species such as hydrogen in the system. The effects of the alpha/beta/gamma radiolysis in relation with the storage conditions were also quantified. The experimental data obtained on spent fuel indicate that the alteration rate of the matrix based on the behaviour of tracer elements (caesium and strontium) reached a maximum value of some mg.m -2 .d -1 , even under very oxidizing conditions. The solubility of uranium and the nature of the secondary phases depend however on the extent of the oxidizing conditions. (author)

Characterization of radioactive contaminants and water treatment trials for the Taiwan Research Reactor's spent fuel pool

International Nuclear Information System (INIS)

Huang, Chun-Ping; Lin, Tzung-Yi; Chiao, Ling-Huan; Chen, Hong-Bin

2012-01-01

Highlights: ► Deal with a practical radioactive contamination in Taiwan Research Reactor spent fuel pool water. ► Identify the properties of radioactive contaminants and performance test for water treatment materials. ► The radioactive solids were primary attributed by ruptured spent fuels, spent resins, and metal debris. ► The radioactive ions were major composed by uranium and fission products. ► Diatomite-based ceramic depth filter can simultaneously removal radioactive solids and ions. - Abstract: There were approximately 926 m 3 of water contaminated by fission products and actinides in the Taiwan Research Reactor's spent fuel pool (TRR SFP). The solid and ionic contaminants were thoroughly characterized using radiochemical analyses, scanning electron microscopy equipped with an energy dispersive spectrometer (SEM-EDS), and inductively coupled plasma optical emission spectrometry (ICP-OES) in this study. The sludge was made up of agglomerates contaminated by spent fuel particles. Suspended solids from spent ion-exchange resins interfered with the clarity of the water. In addition, the ionic radionuclides such as 137 Cs, 90 Sr, U, and α-emitters, present in the water were measured. Various filters and cation-exchange resins were employed for water treatment trials, and the results indicated that the solid and ionic contaminants could be effectively removed through the use of <0.9 μm filters and cation exchange resins, respectively. Interestingly, the removal of U was obviously efficient by cation exchange resin, and the ceramic depth filter composed of diatomite exhibited the properties of both filtration and adsorption. It was found that the ceramic depth filter could adsorb β-emitters, α-emitters, and uranium ions. The diatomite-based ceramic depth filter was able to simultaneously eliminate particles and adsorb ionic radionuclides from water.

Spent fuel handling system for a geologic storage test at the Nevada Test Site

International Nuclear Information System (INIS)

Duncan, J.E.; House, P.A.; Wright, G.W.

1980-01-01

The Lawrence Livermore Laboratory is conducting a test of the geologic storage of encapsulated spent commercial reactor fuel assemblies in a granitic rock at the Nevada Test Site. The test, known as the Spent Fuel Test-Climax (SFT-C), is sponsored by the US Department of Energy, Nevada Operations Office. Eleven pressurized-water-reactor spent fuel assemblies are stored retrievably for three to five years in a linear array in the Climax stock at a depth of 420 m

Strength and water-tightness of the closure head and valves of a model cask under high external pressure

International Nuclear Information System (INIS)

Terada, O.; Kumada, M.; Hayakawa, T.; Mochizuki, S.; Ohrui, K.

1978-01-01

This paper describes experimental research on the strength and water-tightness of the closure head and attached valves of a model cask under high external pressure, in simulation of its having been accidentally lost in the deep sea. Both the external pressure tests and the corrosion tests were carried out using scale models of the closure head of an 80-ton spent-fuel shipping cask, and the full size pressure relief valves and drain valves which were to be attached to the actual cask. Based on the results of the above tests, evaluations were made, and new information was obtained on the pressure-proof strength and water-tightness of the closure head of the cask and the valves. Lastly, research which is being carried on in Japan on the pressure equalizer is also introduced

Corrosion of research reactor Al-clad spent fuel in water

International Nuclear Information System (INIS)

Bendereskaya, O.S.; De, P.K.; Haddad, R.; Howell, J.P.; Johnson, A.B. Jr.; Laoharojanaphand, S.; Luo, S.; Ramanathan, L.V.; Ritchie, I.; Hussain, N.; Vidowsky, I.; Yakovlev, V.

2002-01-01

A significant amount of aluminium-clad spent nuclear fuel from research and test reactors worldwide is currently being stored in water-filled basins while awaiting final disposition. As a result of corrosion issues, which developed from the long-term wet storage of aluminium-clad fuel, the International Atomic Energy Agency (IAEA) implemented a Co-ordinated Research Project (CRP) in 1996 on the 'Corrosion of Research Reactor Aluminium-Clad Spent Fuel in Water'. The investigations undertaken during the CRP involved ten institutes in nine different countries. The IAEA furnished corrosion surveillance racks with aluminium alloys generally used in the manufacture of the nuclear fuel cladding. The individual countries supplemented these racks with additional racks and coupons specific to materials in their storage basins. The racks were immersed in late 1996 in the storage basins with a wide range of water parameters, and the corrosion was monitored at periodic intervals. Results of these early observations were reported after 18 months at the second research co-ordination meeting (RCM) in Sao Paulo, Brazil. Pitting and crevice corrosion were the main forms of corrosion observed. Corrosion caused by deposition of iron and other particles on the coupon surfaces was also observed. Galvanic corrosion of stainless steel/aluminium coupled coupons and pitting corrosion caused by particle deposition was observed. Additional corrosion racks were provided to the CRP participants at the second RCM and were immersed in the individual basins by mid-1998. As in the first set of tests, water quality proved to be the key factor in controlling corrosion. The results from the second set of tests were presented at the third and final RCM held in Bangkok, Thailand in October 2000. An IAEA document giving details about this CRP and other guidelines for spent fuel storage is in pres. This paper presents some details about the CRP and the basis for its extension. (author)

Nuclide inventories of spent fuels from light water reactors

International Nuclear Information System (INIS)

Okumura, Keisuke; Okamoto, Tsutomu

2012-02-01

Accurate information on nuclide inventories of spent fuels from Light Water Reactors (LWRs) is important for evaluations of criticality, decay heat, radioactivity, toxicity, and so on, in the safety assessments of storage, transportation, reprocessing and waste disposal of the spent fuels. So, a lot of lattice burn-up calculations were carried out for the possible fuel specifications and irradiation conditions in Japanese commercial LWRs by using the latest nuclear data library JENDL-4.0 and a sophisticated lattice burn-up calculation code MOSRA-SRAC. As a result, burn-up changes of nuclide inventories and their possible ranges were clarified for 21 heavy nuclides and 118 fission products, which are important from the viewpoint of impacts to nuclear characteristics and nuclear fuel cycle and environment. (author)

CIRFT Data Update and Data Analyses for Spent Nuclear Fuel Vibration Reliability Study

Energy Technology Data Exchange (ETDEWEB)

Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-08-01

The objective of this research is to collect experimental data on spent nuclear fuel (SNF) from pressurized water reactors (PWRs), including the H. B. Robinson Nuclear Power Station (HBR), Catawba Nuclear Station, North Anna Nuclear Power Station (NA), and the Limerick Nuclear Power Station (LMK) boiling water reactor (BWR).

Economics of water basin storage of spent light water reactor fuel

International Nuclear Information System (INIS)

Driggers, F.E.

1978-01-01

As part of the International Spent Fuel Storage program, a preliminary Venture Guidance Assessment of the cost was made. The escalated cost of a reference facility with a capacity to receive 2000 MT/y of spent LWR fuel and to store 5000 MT in water-filled pools was converted to $180 million in 1978 dollars for a stand-alone facility. It was estimated that the receiving rate could be increased to 3000 MT/y for an additional $15 million and that increments could be added to the storage capacity for $13 million per 1000 MT. If a receipt rate of more than 3000 MT/y is required, a new facility in another part of the country might be built to reduce total costs including transportation. Operating costs are determined by the number of people employed and by the costs of stainless steel baskets. An operating crew of 150 is required for the reference facility; the associated cost, including overhead and supplies, is $6 million. During an extended storage-only period, this cost is assumed to drop to $4 million. Fuel baskets are estimated to cost $6.20/kg of spent fuel averaged over a reactor mix of two-thirds PWRs and one-third BWRs. The nominal basket requirements of $10 million for the first year are capitalized. If the facility is financed by the government and a one-time fee is charged to recover all of the away-from-reactor (AFR) basin costs, the fee is about $60/kg of spent fuel plus any government surcharge to cover research and development, overhead, and additional contingencies. If the facility is financed by industry with an annual charge that includes a fixed charge on capital of 25%, the annual fee is about $16/kg-y. In calculating both fees, it is assumed that each storage position is occupied for ten years. 8 tables

Spent fuel element storage facility

International Nuclear Information System (INIS)

Ukaji, Hideo; Yamashita, Rikuo.

1981-01-01

Purpose: To always keep water level of a spent fuel cask pit equal with water level of spent fuel storage pool by means of syphon principle. Constitution: The pool water of a spent fuel storage pool is airtightly communicated through a pipe with the pool water of a spent fuel cask, and a gate is provided between the pool and the cask. Since cask is conveyed into the cask pit as the gate close while conveying, the pool water level is raised an amount corresponding to the volume of the cask, and water flow through scattering pipe and the communication pipe to the storage pool. When the fuel is conveyed out of the cask, the water level is lowered in the amount corresponding to the volume in the cask pit, and the water in the pool flow through the communication pipe to the cask pit. (Sekiya, K.)

Time Spent on the Internet and Adolescent Blood Pressure.

Science.gov (United States)

Cassidy-Bushrow, Andrea E; Johnson, Dayna A; Peters, Rosalind M; Burmeister, Charlotte; Joseph, Christine L M

2015-10-01

Internet use is nearly ubiquitous among adolescents. Growing evidence suggests heavy Internet use negatively impacts health, yet the relationship between time spent on the Internet and adolescent blood pressure (BP) is unknown. We examined the association between Internet use and elevated BP in a racially diverse cross-sectional sample of 331 healthy adolescents (ages 14-17 years). Heavy Internet use was defined as ≥ 2 hr/day, moderate use as Internet users had statistically significantly higher odds of elevated BP compared to light Internet users. School nurses can play an important role in preventing high BP through assessment of BP and other health behaviors including Internet use, and health teaching to individuals, student groups, faculty, and parents to increase awareness of the relationship between Internet use and health. © The Author(s) 2014.

Radionuclide distribution in LWR [light-water reactor] spent fuel

International Nuclear Information System (INIS)

Guenther, R.J.; Blahnik, D.E.; Thomas, L.E.; Baldwin, D.L.; Mendel, J.E.

1990-09-01

The Materials Characterization Center (MCC) at Pacific Northwest Laboratory (PNL) provides well-characterized spent fuel from light-water reactors (LWRs) for use in laboratory tests relevant to nuclear waste disposal in the proposed Yucca Mountain repository. Interpretation of results from tests on spent fuel oxidation, dissolution, and cladding degradation requires information on the inventory and distribution of radionuclides in the initial test materials. The MCC is obtaining this information from examinations of Approved Testing Materials (ATMs), which include spent fuel with burnups from 17 to 50 MWd/kgM and fission gas releases (FGR) from 0.2 to 18%. The concentration and distribution of activation products and the release of volatile fission products to the pellet-cladding gap and rod plenum are of particular interest because these characteristics are not well understood. This paper summarizes results that help define the 14 C inventory and distribution in cladding, the ''gap and grain boundary'' inventory of radionuclides in fuels with different FGRs, and the structure and radionuclide inventory of the fuel rim region within a few hundred micrometers from the fuel edge. 6 refs., 5 figs., 1 tab

The Proliferation Resistance of a Nuclear Fuel Cycle Using Fuel Recovered from the Electrolytic Reduction of Pressurized Water Reactor Spent Fuel

Energy Technology Data Exchange (ETDEWEB)

Kang, Jung Min; Cochran, Thomas; Mckinzie, Matthew [NRDC, Washington, (United States)

2016-05-15

At some points in the fuel cycle, a level of intrinsic or technical proliferation-resistance can be provided by radiation barriers that surround weapons-usable materials. In this report we examine some aspects of intrinsic proliferation resistance of a fuel cycle for a fast neutron reactor that uses fuel recovered from the electrolytic reduction process of pressurized water reactor spent fuel, followed by a melt-refining process. This fuel cycle, proposed by a nuclear engineer at the Korea Advanced Institute of Science and Technology (KAIST), is being examined with respect to its potential merits of higher fuel utilization, lower production of radioactive byproducts, and better economics relative to a pyroprocesing-based fuel cycle. With respect to intrinsic proliferation resistance, however, we show that since europium is separated out during the electrolytic reduction process, this fuel cycle has little merit beyond that of a pyroprocessing-based fuel cycle because of the lower radiation barrier of its recovered materials containing weapons-usable actinides. Unless europium is not separated following voloxidation, the proposed KAIST fuel cycle is not intrinsically proliferation resistant and in this regard does not represent a significant improvement over pyroprocessing. We suggest further modification of the proposed KAIST fuel cycle, namely, omitting electrolytic reduction and melt reduction, and producing the fast reactor fuel directly following voloxidation.

The isolation of water-soluble radionuclides from deteriorating spent nuclear fuel in zeolite cartridge

International Nuclear Information System (INIS)

Hassan, N.M.; Thompson, M.C.

1996-01-01

A method of isolating water-soluble radionuclides leaching from deteriorating spent nuclear fuel by ion-exchange in zeolite cartridges has been studied. Design calculations of two zeolite cartridges to be incorporated in typical spent fuel storage bundle have been provided. Equilibrium exchange data obtained at several temperatures have shown that the maximum exchange capacity of total cesium in sodium titanium aluminosilicate was 114 mg/g zeolite and the capacity at 95% exchange for radioactive isotope Cs-137 was calculated as 55.2 mg/g zeolite. The kinetic data suggest that the rate of exchange of Cs + in sodium titanium aluminosilicate zeolite takes place by a fast initial exchange step followed by slow diffusion of cesium cations. Design calculations based on the equilibrium exchange data show that water-soluble radionuclides leaching from Mk 31 slugs can be isolated using two zeolite cartridges, each 3.7 inches in inside diameter and 2.5 inches in length. The cartridges are designed to isolate 95% of the Cs + leaching from the spent fuel storage bundle. The results from the thermal induced convective flow tests indicate that the system will provide necessary cooling to the spent fuel by convective currents while isolating the Cs + leaching from spent fuel storage bundle in the cartridges

Current state of knowledge of water radiolysis effects on spent nuclear fuel corrosion

International Nuclear Information System (INIS)

Christensen, H.; Sunder, S.

2000-07-01

Literature data on the effect of water radiolysis products on spent-fuel oxidation and dissolution are reviewed. Effects of gamma radiolysis, alpha radiolysis, and dissolved O 2 or H 2 O 2 in unirradiated solutions are discussed separately. Also, the effect of carbonate in gamma-irradiated solutions and radiolysis effects on leaching of spent fuel are reviewed. In addition, a kinetic model for calculating the corrosion rates of UO 2 in solutions undergoing radiolysis is discussed. The model gives good agreement between calculated and measured corrosion rates in the case of gamma radiolysis and in unirradiated solutions containing dissolved oxygen or hydrogen peroxide. However, the model fails to predict the results of alpha radiolysis. In a recent study , it was shown that the model gave good agreement with measured corrosion rates of spent fuel exposed in deionized water. The applications of radiolysis studies for geologic disposal of used nuclear fuel are discussed. (author)

Spent fuel workshop'2002

International Nuclear Information System (INIS)

Poinssot, Ch.

2002-01-01

This document gathers the transparencies of the presentations given at the 2002 spent fuel workshop: Session 1 - Research Projects: Overview on the IN CAN PROCESSES European project (M. Cowper), Overview on the SPENT FUEL STABILITY European project (C. Poinssot), Overview on the French R and D project on spent fuel long term evolution, PRECCI (C. Poinssot); Session 2 - Spent Fuel Oxidation: Oxidation of uranium dioxide single crystals (F. Garrido), Experimental results on SF oxidation and new modeling approach (L. Desgranges), LWR spent fuel oxidation - effects of burn-up and humidity (B. Hanson), An approach to modeling CANDU fuel oxidation under dry storage conditions (P. Taylor); Session 3 - Spent Fuel Dissolution Experiments: Overview on high burnup spent fuel dissolution studies at FZK/INE (A. Loida), Results on the influence of hydrogen on spent fuel leaching (K. Spahiu), Leaching of spent UO 2 fuel under inert and reducing conditions (Y. Albinsson), Fuel corrosion investigation by electrochemical techniques (D. Wegen), A reanalysis of LWR spent fuel flow through dissolution tests (B. Hanson), U-bearing secondary phases formed during fuel corrosion (R. Finch), The near-field chemical conditions and spent fuel leaching (D. Cui), The release of radionuclides from spent fuel in bentonite block (S.S. Kim), Trace actinide behavior in altered spent fuel (E. Buck, B. Hanson); Session 4 - Radiolysis Issues: The effect of radiolysis on UO 2 dissolution determined from electrochemical experiments with 238 Pu doped UO 2 M. Stroess-Gascoyne (F. King, J.S. Betteridge, F. Garisto), doped UO 2 studies (V. Rondinella), Preliminary results of static and dynamic dissolution tests with α doped UO 2 in Boom clay conditions (K. Lemmens), Studies of the behavior of UO 2 / water interfaces under He 2+ beam (C. Corbel), Alpha and gamma radiolysis effects on UO 2 alteration in water (C. Jegou), Behavior of Pu-doped pellets in brines (M. Kelm), On the potential catalytic behavior of

Good Practices for Water Quality Management in Research Reactors and Spent Fuel Storage Facilities

International Nuclear Information System (INIS)

2011-01-01

Water is the most common fluid used to remove the heat produced in a research reactor (RR). It is also the most common media used to store spent fuel elements after being removed from the reactor core. Spent fuel is stored either in the at-reactor pool or in away-from-reactor wet facilities, where the fuel elements are maintained until submission to final disposal, or until the decay heat is low enough to allow migration to a dry storage facility. Maintaining high quality water is the most important factor in preventing degradation of aluminium clad fuel elements, and other structural components in water cooled research reactors. Excellent water quality in spent fuel wet storage facilities is essential to achieve optimum storage performance. Experience shows the remarkable success of many research reactors where the water chemistry has been well controlled. In these cases, aluminium clad fuel elements and aluminium pool liners show few, if any, signs of either localized or general corrosion, even after more than 30 years of exposure to research reactor water. In contrast, when water quality was allowed to degrade, the fuel clad and the structural parts of the reactor have been seriously corroded. The driving force to prepare this publication was the recognition that, even though a great deal of information on research reactor water quality is available in the open literature, no comprehensive report addressing the rationale of water quality management in research reactors has been published to date. This report is designed to provide a comprehensive catalogue of good practices for the management of water quality in research reactors. It also presents a brief description of the corrosion process that affects the components of a research reactor. Further, the report provides a basic understanding of water chemistry and its influence on the corrosion process; specifies requirements and operational limits for water purification systems of RRs; describes good practices

Nuclear criticality safety studies applicable to spent fuel shipping cask designs and spent fuel storage

International Nuclear Information System (INIS)

Tang, J.S.

1980-11-01

Criticality analyses of water-moderated and reflected arrays of LWR fresh and spent fuel assemblies were carried out in this study. The calculated results indicate that using the assumption of fresh fuel loading in spent fuel shipping cask design leads to assembly spacings which are about twice the spacings of spent fuel loadings. Some shipping cask walls of composite lead and water are more effective neutron reflectors than water of 30.48 cm

Water activities in Forsmark (Part II). The final disposal facility for spent fuel: water activities above ground

International Nuclear Information System (INIS)

Werner, Kent; Hamren, Ulrika; Collinder, Per; Ridderstolpe, Peter

2010-09-01

The construction of the repository for spent nuclear fuel in Forsmark is associated with a number of measures above ground that constitute water operations according to Chapter 11 in the Swedish Environmental Code. This report, which is an appendix to the Environmental Impact Assessment, describes these water operations, their effects and consequences, and planned measures

Determination of maximum water temperature within the spent fuel pool of Angra Nuclear Power Plant - Unit 3

Energy Technology Data Exchange (ETDEWEB)

Werner, F.L., E-mail: fernanda.werner@poli.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Departamento de Engenharia Nuclear; Alves, A.S.M., E-mail: asergi@eletronuclear.gov.br [Eletrobras Termonuclear (Eletronuclear), Rio de Janeiro, RJ (Brazil); Frutuoso e Melo, P.F., E-mail: frutuoso@nuclear.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil)

2017-07-01

In this paper, a mathematical model for the determination of the maximum water temperature within the spent fuel pool of Angra Nuclear Power Plant – Unit 3 was developed. The model was obtained from the boundary layer analysis and the application of Navier-Stokes equation to a vertical flat plate immersed in a water flow under free convection regime. Both types of pressure loss coefficients through the flow channel were considers in the modeling, the form coefficient for fuel assemblies (FAs) and the loss due to rod friction. The resulting equations enabled the determination of a mixed water temperature below the storage racks (High Density Storage Racks) as well as the estimation of a temperature gradient through the racks. The model was applied to the authorized operation of the plant (power operation, plant outage and upset condition) and faulted conditions (loss of coolant accidents and external events). The results obtained are in agreement with Brazilian and international standards. (author)

Determination of maximum water temperature within the spent fuel pool of Angra Nuclear Power Plant - Unit 3

International Nuclear Information System (INIS)

Werner, F.L.; Frutuoso e Melo, P.F.

2017-01-01

In this paper, a mathematical model for the determination of the maximum water temperature within the spent fuel pool of Angra Nuclear Power Plant – Unit 3 was developed. The model was obtained from the boundary layer analysis and the application of Navier-Stokes equation to a vertical flat plate immersed in a water flow under free convection regime. Both types of pressure loss coefficients through the flow channel were considers in the modeling, the form coefficient for fuel assemblies (FAs) and the loss due to rod friction. The resulting equations enabled the determination of a mixed water temperature below the storage racks (High Density Storage Racks) as well as the estimation of a temperature gradient through the racks. The model was applied to the authorized operation of the plant (power operation, plant outage and upset condition) and faulted conditions (loss of coolant accidents and external events). The results obtained are in agreement with Brazilian and international standards. (author)

Characterization of radioactive contaminants and water treatment trials for the Taiwan Research Reactor's spent fuel pool

Energy Technology Data Exchange (ETDEWEB)

Huang, Chun-Ping, E-mail: chunping@iner.gov.tw [Institute of Nuclear Energy Research, 1000, Wenhua Road, Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan, ROC (China); Lin, Tzung-Yi; Chiao, Ling-Huan; Chen, Hong-Bin [Institute of Nuclear Energy Research, 1000, Wenhua Road, Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan, ROC (China)

2012-09-30

Highlights: Black-Right-Pointing-Pointer Deal with a practical radioactive contamination in Taiwan Research Reactor spent fuel pool water. Black-Right-Pointing-Pointer Identify the properties of radioactive contaminants and performance test for water treatment materials. Black-Right-Pointing-Pointer The radioactive solids were primary attributed by ruptured spent fuels, spent resins, and metal debris. Black-Right-Pointing-Pointer The radioactive ions were major composed by uranium and fission products. Black-Right-Pointing-Pointer Diatomite-based ceramic depth filter can simultaneously removal radioactive solids and ions. - Abstract: There were approximately 926 m{sup 3} of water contaminated by fission products and actinides in the Taiwan Research Reactor's spent fuel pool (TRR SFP). The solid and ionic contaminants were thoroughly characterized using radiochemical analyses, scanning electron microscopy equipped with an energy dispersive spectrometer (SEM-EDS), and inductively coupled plasma optical emission spectrometry (ICP-OES) in this study. The sludge was made up of agglomerates contaminated by spent fuel particles. Suspended solids from spent ion-exchange resins interfered with the clarity of the water. In addition, the ionic radionuclides such as {sup 137}Cs, {sup 90}Sr, U, and {alpha}-emitters, present in the water were measured. Various filters and cation-exchange resins were employed for water treatment trials, and the results indicated that the solid and ionic contaminants could be effectively removed through the use of <0.9 {mu}m filters and cation exchange resins, respectively. Interestingly, the removal of U was obviously efficient by cation exchange resin, and the ceramic depth filter composed of diatomite exhibited the properties of both filtration and adsorption. It was found that the ceramic depth filter could adsorb {beta}-emitters, {alpha}-emitters, and uranium ions. The diatomite-based ceramic depth filter was able to simultaneously

Normal and compact spent fuel storage in light water reactor power plants

International Nuclear Information System (INIS)

Kuenel, R.R.

1978-01-01

The compact storage of light water reactor spent fuel is a safe, cheap and reliable contribution towards overcoming the momentarily existing shortage in spent fuel reprocessing. The technical concept is described and physical behaviour discussed. The introduction of compact storage racks in nuclear power plants increases the capacity from 100 to about 240 %. The increase in decay heat is not more than about 14%, the increase in activity inventory and hazard potential does not exceed 20%. In most cases the existing power plant equipment fulfils the new requirements. (author)

Corrosion surveillance programme for Latin American research reactor Al-clad spent fuel in water

International Nuclear Information System (INIS)

Ramanathan, L.V.; Haddad, R.; Ritchie, I.

2002-01-01

The objectives of the IAEA sponsored Regional Technical Co-operation Project for Latin America (Argentina, Brazil, Chile, Mexico, and Peru) are to provide the basic conditions to define a regional strategy for managing spent fuel and to provide solutions, taking into consideration the economic and technological realities of the countries involved. In particular, to determine the basic conditions for managing research reactor spent fuel during operation and interim storage as well as final disposal, and to establish forms of regional cooperation in the four main areas: spent fuel characterization, safety, regulation and public communication. This paper reports the corrosion surveillance activities of the Regional Project and these are based on the IAEA sponsored co-ordinated research project (CRP) on 'Corrosion of research reactor Al-clad spent fuel in water'. The overall test consists of exposing corrosion coupon racks at different spent fuel basins followed by evaluation. (author)

Spent fuel dissolution rates as a function of burnup and water chemistry

International Nuclear Information System (INIS)

Gray, W.J.

1998-06-01

To help provide a source term for performance-assessment calculations, dissolution studies on light-water-reactor (LWR) spent fuel have been conducted over the past few years at Pacific Northwest National Laboratory in support of the Yucca Mountain Site Characterization Project. This report describes that work for fiscal years 1996 through mid-1998 and includes summaries of some results from previous years for completeness. The following conclusions were based on the results of various flowthrough dissolution rate tests and on tests designed to measure the inventories of 129 I located within the fuel/cladding gap region of different spent fuels: (1) Spent fuels with burnups in the range 30 to 50 MWd/kgM all dissolved at about the same rate over the conditions tested. To help determine whether the lack of burnup dependence extends to higher and lower values, tests are in progress or planned for spent fuels with burnups of 13 and ∼ 65 MWd/kgM. (2) Oxidation of spent fuel up to the U 4 O 9+x stage does not have a large effect on intrinsic dissolution rates. However, this degree of oxidation could increase the dissolution rates of relatively intact fuel by opening the grain boundaries, thereby increasing the effective surface area that is available for contact by water. From a disposal viewpoint, this is a potentially more important consideration than the effect on intrinsic rates. (3) The gap inventories of 129 I were found to be smaller than the fission gas release (FGR) for the same fuel rod with the exception of the rod with the highest FGR. Several additional fuels would have to be tested to determine whether a generalized relationship exists between FGR and 129 I gap inventory for US LWR fuels

Performance Variation of Spent Resin in Mixed Bed From Water Purifying System of Xi'an Pulse Reactor

International Nuclear Information System (INIS)

Li Hua; Ma Yan; Xiao Yan; Liu Yueheng; Yang Yongqing

2010-01-01

Detailed physical and chemical characteristic analysis was performed on the spent cation and anion resins in the mixed bed from Xi'an Pulse Reactor water purifying system.The exchange performance variations of used resins and the contributions from different factors to the variation were discussed.Based on the obtained information of the impurities in the used resin, the contamination state of the water in the Xi'an Pulse Reactor water pool, the corrosion state of the structural material in the reactor was presented. The spent anion resin almost completely losses its exchange performance,while the remaining exchange capacity in the spent cation resin is still high.The radiation field from the reactor operation contributes little to the degradation of the performance of the resins. The exchange capacity loss of the spent anion resin is due to the exchange of its active groups into abundant carbonate and a certain amount of organics. The impurity amount in the anion and cation exchange resins is low,which suggests(that) the water in the Xi'an Pulse Reactor water pool is little contaminated. A certain extent of corrosion is occurred on the structural material in the swimming pool of the reactor. The results provide important referential data for the operational safety of the water purifying system of similar research reactor. (authors)

Possibilities for the efficient utilisation of spent geothermal waters.

Science.gov (United States)

Tomaszewska, Barbara; Szczepański, Andrzej

2014-10-01

Waters located at greater depths usually exhibit high mineral content, which necessitates the use of closed systems, i.e. re-injecting them into the formation after recovering the heat. This significantly reduces investment efficiency owing to the need to drill absorption wells and to perform anti-corrosion and anti-clogging procedures. In this paper, possibilities for the efficient utilisation of cooled geothermal waters are considered, particularly with respect to open or mixed geothermal water installations. Where cooled water desalination technologies are used, this allows the water to be demineralised and used to meet local needs (as drinking water and for leisure purposes). The retentate left as a by-product of the process contains valuable ingredients that can be used for balneological and/or leisure purposes. Thus, the technology for desalinating spent geothermal waters with high mineral content allows improved water management on a local scale and makes it possible to minimise the environmental threat resulting from the need to dump these waters into waterways or surface water bodies and/or inject them into the formation. The paper is concerned with Polish geothermal system and provides information about the parameters of Polish geothermal waters.

Corrosion of research reactor aluminium clad spent fuel in water

International Nuclear Information System (INIS)

2003-01-01

This report describes research performed in ten laboratories within the framework of the IAEA Co-ordinated Research Project on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water. The project consisted of exposure of standard racks of corrosion coupons in the spent fuel pools of the participating research reactor laboratories and the evaluation of the coupons after predetermined exposure times, along with periodic monitoring of the storage water. A group of experts in the field contributed a state of the art review and provided technical supervision of the project. Localized corrosion mechanisms are notoriously difficult to understand, and it was clear from the outset that obtaining consistency in the results and their interpretation from laboratory to laboratory would depend on the development of an excellent set of experimental protocols. These experimental protocols are described in the report together with guidelines for the maintenance of optimum water chemistry to minimize the corrosion of aluminium clad research reactor fuel in wet storage. A large database on corrosion of aluminium clad materials has been generated from the CRP and the SRS corrosion surveillance programme. An evaluation of these data indicates that the most important factors contributing to the corrosion of the aluminium are: (1) High water conductivity (100-200 μS/cm); (2) Aggressive impurity ion concentrations (Cl - ); (3) Deposition of cathodic particles on aluminium (Fe, etc.); (4) Sludge (containing Fe, Cl - and other ions in concentrations greater than ten times the concentrations in the water); (5) Galvanic couples between dissimilar metals (stainless steel-aluminium, aluminium-uranium, etc); (6) Scratches and imperfections (in protective oxide coating on cladding); (7) Poor water circulation. These factors operating both independently and synergistically may cause corrosion of the aluminium. The single most important key to preventing corrosion is maintaining good

International Atomic Energy Agency (IAEA) Activity on Technical Influence of High Burnup UOX and MOX Water Reactor Fuel on Spent Fuel Management

International Nuclear Information System (INIS)

Lovasic, Z.; Einziger, R.

2009-01-01

This paper briefly reviews the results of the International Atomic Energy Agency (IAEA) project investigating the influence of high burnup and mixed-oxide (MOX) fuels, from water power reactors, on spent fuel management. These data will provide information on the impacts, regarding spent fuel management, for those countries operating light-water reactors (LWR)s and heavy-water reactors (HWR)s with zirconium alloy-clad uranium dioxide (UOX) fuels, that are considering the use of higher burnup UOX or the introduction of reprocessing and MOX fuels. The mechanical designs of lower burnup UOX and higher burnup UOX or MOX fuel are very similar, but some of the properties (e.g., higher fuel rod internal pressures; higher decay heat; higher specific activity; and degraded cladding mechanical properties of higher burnup UOX and MOX spent fuels) may potentially significantly affect the behavior of the fuel after irradiation. These properties are reviewed. The effects of these property changes on wet and dry storage, transportation, reprocessing, re-fabrication of fuel, and final disposal were evaluated, based on regulatory, safety, and operational considerations. Political and strategic considerations were not taken into account since relative importance of technical, economic and strategic considerations vary from country to country. There will also be an impact of these fuels on issues like non-proliferation, safeguards, and sustainability, but because of the complexity of factors affecting those issues, they are only briefly discussed. Data gaps were also identified during this investigation. The pros and cons of using high burnup UOX or MOX, for each applicable issue in each stage of the back end of the fuel cycle, were evaluated and are discussed.. Although, in theory, higher burnup fuel and MOX fuels mean a smaller quantity of spent fuel, the potential need for some changes in design of spent fuel storage, transportation, handling, reprocessing, re-fabrication, and

BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

Energy Technology Data Exchange (ETDEWEB)

Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mertyurek, Ugur [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-10-01

The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is

Investigation of water-logged spent fuel rods under dry storage conditions

International Nuclear Information System (INIS)

Kohli, R.; Pasupathi, V.

1986-09-01

Tests were conducted to determine the amount of moisture contained in breached, water-logged spent fuel rods and the rate of release. Two well-characterized BWR fuel rods with reactor-induced breaches were tested in a hot cell. These rods contained approximately 6 to 10 g of moisture, most of which was released during heating tests simulating normal cask drying operations. Additional testing with two intentionally defected fuel rods (BWR and PWR) was performed to evaluate the effect of the cladding breach on migration of moisture along the length of the fuel rod. The results showed that the moisture released from reactor-breached spent fuel rods was insufficient to cause degradation of fuel or dry storage system components

Management of Spent Nuclear Fuel from Nuclear Power Plant Reactor

International Nuclear Information System (INIS)

Wati, Nurokhim

2008-01-01

Management of spent nuclear fuel from Nuclear Power Plant (NPP) reactor had been studied to anticipate program of NPP operation in Indonesia. In this paper the quantity of generated spent nuclear fuel (SNF) is predicted based on the national electrical demand, power grade and type of reactor. Data was estimated using Pressurized Water Reactor (PWR) NPP type 1.000 MWe and the SNF management overview base on the experiences of some countries that have NPP. There are four strategy nuclear fuel cycle which can be developed i.e: direct disposal, reprocessing, DUPlC (Direct Use of Spent PWR Fuel In Candu) and wait and see. There are four alternative for SNF management i.e : storage at the reactor building (AR), away from reactor (AFR) using wet centralized storage, dry centralized storage AFR and prepare for reprocessing facility. For the Indonesian case, centralized facility of the wet type is recommended for PWR or BWR spent fuel. (author)

Spent fuel workshop'2002

Energy Technology Data Exchange (ETDEWEB)

Poinssot, Ch

2002-07-01

This document gathers the transparencies of the presentations given at the 2002 spent fuel workshop: Session 1 - Research Projects: Overview on the IN CAN PROCESSES European project (M. Cowper), Overview on the SPENT FUEL STABILITY European project (C. Poinssot), Overview on the French R and D project on spent fuel long term evolution, PRECCI (C. Poinssot); Session 2 - Spent Fuel Oxidation: Oxidation of uranium dioxide single crystals (F. Garrido), Experimental results on SF oxidation and new modeling approach (L. Desgranges), LWR spent fuel oxidation - effects of burn-up and humidity (B. Hanson), An approach to modeling CANDU fuel oxidation under dry storage conditions (P. Taylor); Session 3 - Spent Fuel Dissolution Experiments: Overview on high burnup spent fuel dissolution studies at FZK/INE (A. Loida), Results on the influence of hydrogen on spent fuel leaching (K. Spahiu), Leaching of spent UO{sub 2} fuel under inert and reducing conditions (Y. Albinsson), Fuel corrosion investigation by electrochemical techniques (D. Wegen), A reanalysis of LWR spent fuel flow through dissolution tests (B. Hanson), U-bearing secondary phases formed during fuel corrosion (R. Finch), The near-field chemical conditions and spent fuel leaching (D. Cui), The release of radionuclides from spent fuel in bentonite block (S.S. Kim), Trace actinide behavior in altered spent fuel (E. Buck, B. Hanson); Session 4 - Radiolysis Issues: The effect of radiolysis on UO{sub 2} dissolution determined from electrochemical experiments with {sup 238}Pu doped UO{sub 2} M. Stroess-Gascoyne (F. King, J.S. Betteridge, F. Garisto), doped UO{sub 2} studies (V. Rondinella), Preliminary results of static and dynamic dissolution tests with {alpha} doped UO{sub 2} in Boom clay conditions (K. Lemmens), Studies of the behavior of UO{sub 2} / water interfaces under He{sup 2+} beam (C. Corbel), Alpha and gamma radiolysis effects on UO{sub 2} alteration in water (C. Jegou), Behavior of Pu-doped pellets in brines

Generic environmental impact statement on handling and storage of spent light water power reactor fuel. Appendices

International Nuclear Information System (INIS)

1978-03-01

Detailed appendices are included with the following titles: light water reactor fuel cycle, present practice, model 1000MW(e) coal-fired power plant, increasing fuel storage capacity, spent fuel transshipment, spent fuel generation and storage data (1976-2000), characteristics of nuclear fuel, and ''away-from-reactor'' storage concept

High Pressure Water Jet System Performance Assessment Project A-2A

International Nuclear Information System (INIS)

FARWICK, C.C.

1999-01-01

Performance assessment for canister cleaning system in the KE Basin. Information obtained from this assessment will be used to design any additional equipment used to clean canisters. After thorough review of the design, maintenance history and operational characteristics of the 105 K East (KE) canister cleaning system, Bartlett recommends that the high pressure water jet system (HPWJS) be modified as outlined in section 5.0, and retained for future use. Further, it is recommended that Spent Nuclear Fuel (SNF) Project consider use of a graded approach for canister cleaning, based on individual canister type and characteristics. This approach would allow a simple method to be used on canisters not needing the more rigorous, high-pressure method. Justification is provided in section 5.0. Although Bartlett has provided some preliminary cost estimates, it is recommended that SNF Project perform a detailed cost-benefit analysis to weigh the alternatives presented

High temperature pressure water's blowdown into water. Experimental results

International Nuclear Information System (INIS)

Ishida, Toshihisa; Kusunoki, Tsuyoshi; Kasahara, Yoshiyuki; Iida, Hiromasa

1994-01-01

The purpose of the present experimental study is to clarify the phenomena in blowdown of high temperature and pressure water in pressure vessel into the containment water for evaluation of design of an advanced marine reactor(MRX). The water blown into the containment water flushed and formed steam jet plume. The steam jet condensed in the water, but some stream penetrated to gas phase of containment and contributed to increase of containment pressure. (author)

Radionuclide release from PWR spent fuel specimens with induced cladding defects

International Nuclear Information System (INIS)

Wilson, C.N.; Oversby, V.M.

1984-03-01

Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented

Formation of water-soluble soybean polysaccharides from spent flakes by hydrogen peroxide treatment

DEFF Research Database (Denmark)

Pierce, Brian; Wichmann, Jesper; Tran, Tam H.

2016-01-01

70% of the original insoluble material as high molar mass soluble polysaccharides. A design of experiment was used to quantify the effects of pH, reaction time, and hydrogen peroxide concentration on the reaction yield, average molar mass, and free monosaccharides generated. The resulting product......In this paper we propose a novel chemical process for the generation of water-soluble polysaccharides from soy spent flake, a by-product of the soy food industry. This process entails treatment of spent flake with hydrogen peroxide at an elevated temperature, resulting in the release of more than...... is low in protein, fat, and minerals and contains predominantly water-soluble polysaccharides of high molar mass, including arabinan, type I arabinogalactan, homogalacturonan, xyloglucan, rhamnogalacturonan, and (glucurono)arabinoxylan. This treatment provides a straightforward approach for generation...

Dataset on the spent filter backwash water treatment by sedimentation, coagulation and ultra filtration

OpenAIRE

Mokhtar Mahdavi; Afshin Ebrahimi; Hossein Azarpira; Hamid Reza Tashauoei; Amir Hossein Mahvi

2017-01-01

During operation of most water treatment plants, spent filter backwash water (SFBW) is generated, which accounts about 2â10% of the total plant production. By increasing world population and water shortage in many countries, SFBW can be used as a permanent water source until the water treatment plant is working. This data article reports the practical method being used for water reuse from SFBW through different method including pre-sedimentation, coagulation and flocculation, second clarific...

Development of a strategy for the management of PBMR spent fuel in South Africa

Energy Technology Data Exchange (ETDEWEB)

Smith, S.W., E-mail: Schalk.Smith@necsa.co.z [South African Nuclear Energy Corporation Ltd (Necsa), Pretoria 0001 (South Africa); Bredell, P.J. [South African Nuclear Energy Corporation Ltd (Necsa), Pretoria 0001 (South Africa)

2010-10-15

South Africa is planning to expand its nuclear power generating capacity by deploying a number of pressurized-water reactors and pebble-bed modular reactors. It can be expected that this program will impact on the current and planned spent fuel and radioactive waste management systems in South Africa. This paper proposes an approach to develop a strategy for the management of PBMR spent fuel that would form an integral part of the overall national radwaste management system. The approach is expected to provide a conceptual spent fuel management strategy and will also highlight areas that need to be further developed, thus providing guidance for basic technology development.

Water chemistry management of the spent-fuel pool in Thailand

International Nuclear Information System (INIS)

Suparit, Nitaya; Sukharn, Sumalee; Busamongkol, Arporn; Laoharojanaphand, Sirinart

1999-01-01

Water chemistry of the OAEP spent-fuel pool has been closely monitored without any pre-treatment for its conductivity, pH, temperature, chloride ion, sulfate ion, nitrate ion, phosphate ion, silver ion, and copper ion as well as its gamma activity of Cs-137. Conductivity, pH and temperature were measured using a portable pH and conductivity meter with built in temperature probe. Chloride ion was monitored by an automatic micro-titrator with silver nitrate as titrant and platinum indicator electrode. Nitrate, sulfate and phosphate were analysed by ion-exchange chromatographic method using an anion separator column and salicylate buffer as eluant. Gamma activity of Cs-137 was measured using a Canberra gamma spectrometer with HpGe detector. Silver and copper were analysed by ICP-AES technique within 6 hours after collection. During the study period from March 1996-September 1998, the conductivity was between l.25-4.80 μ/cm, pH in the range of 5-8.1, and temperature from 26.4-29.6 degree celsius. Chloride ion was found between 0.l-0.8 ppm. Silver, copper, nitrate, sulfate and phosphate ions were undetectable. Overall chemical composition of the water shows that the water is kept in standard condition recommended for safety storage. However, the presence of gamma activity of Cs-137 (average value of 138 Bq/l) indicates a slight leak of the spent fuel. (author)

Water-Based Pressure-Sensitive Paints

Science.gov (United States)

Jordan, Jeffrey D.; Watkins, A. Neal; Oglesby, Donald M.; Ingram, JoAnne L.

2006-01-01

Water-based pressure-sensitive paints (PSPs) have been invented as alternatives to conventional organic-solvent-based pressure-sensitive paints, which are used primarily for indicating distributions of air pressure on wind-tunnel models. Typically, PSPs are sprayed onto aerodynamic models after they have been mounted in wind tunnels. When conventional organic-solvent-based PSPs are used, this practice creates a problem of removing toxic fumes from inside the wind tunnels. The use of water-based PSPs eliminates this problem. The waterbased PSPs offer high performance as pressure indicators, plus all the advantages of common water-based paints (low toxicity, low concentrations of volatile organic compounds, and easy cleanup by use of water).

Burn-up credit in criticality safety of PWR spent fuel

Energy Technology Data Exchange (ETDEWEB)

Mahmoud, Rowayda F., E-mail: Rowayda_mahmoud@yahoo.com [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Shaat, Mohamed K. [Nuclear Engineering, Reactors Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Nagy, M.E.; Agamy, S.A. [Professor of Nuclear Engineering, Nuclear and Radiation Department, Alexandria University (Egypt); Abdelrahman, Adel A. [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt)

2014-12-15

Highlights: • Designing spent fuel wet storage using WIMS-5D and MCNP-5 code. • Studying fresh and burned fuel with/out absorber like “B{sub 4}C and Ag–In–Cd” in racks. • Sub-criticality was confirmed for fresh and burned fuel under specific cases. • Studies for BU credit recommend increasing fuel burn-up to 60.0 GWD/MTU. • Those studies require new core structure materials, fuel composition and cladding. - Abstract: The criticality safety calculations were performed for a proposed design of a wet spent fuel storage pool. This pool will be used for the storage of spent fuel discharged from a typical pressurized water reactor (PWR). The mathematical model based on the international validated codes, WIMS-5 and MCNP-5 were used for calculating the effective multiplication factor, k{sub eff}, for the spent fuel stored in the pool. The data library for the multi-group neutron microscopic cross-sections was used for the cell calculations. The k{sub eff} was calculated for several changes in water density, water level, assembly pitch and burn-up with different initial fuel enrichment and new types and amounts of fixed absorbers. Also, k{sub eff} was calculated for the conservative fresh fuel case. The results of the calculations confirmed that the effective multiplication factor for the spent fuel storage is sub-critical for all normal and abnormal states. The future strategy for the burn-up credit recommends increasing the fuel burn-up to a value >60.0 GWD/MTU, which requires new fuel composition and new fuel cladding material with the assessment of the effects of negative reactivity build up.

Safety of spent fuel elements storage under water at La Hague facility

International Nuclear Information System (INIS)

Guezenec, J.Y.

1990-12-01

Awaiting for a decision about radioactive waste repository, the spent fuel elements are stored in the storage pools at the La Hague facility. The water in the pools is permanently cooled and purified to maintain the temperature, radioactivity and chemical pollution under preset limits. The first safety problem is concerned with the spent fuel transport casks. Opening of the casks is done under water in a number of facilities. The most recent approach is done by the company To, which established dry manipulation which enables to minimise the risk of possible cask failures as well as external contamination of cooling fins of the casks. Another general safety related problem is related to criticality risk caused by possible cooling failures or by external events like earthquakes. Special probability limit is set up for seismic events to be less than 10 -7 /year. Equally, risk of fuel assembly failures due to possible chocs and possibility of defects in pool isolation are taken into account [fr

Near surface spent fuel storage: environmental issues

International Nuclear Information System (INIS)

Nelson, I.C.; Shipler, D.B.; McKee, R.W.; Glenn, R.D.

1979-01-01

Interim storage of spent fuel appears inevitable because of the lack of reprocessing plants and spent fuel repositories. This paper examines the environmental issues potentially associated with management of spent fuel before disposal or reprocessing in a reference scenario. The radiological impacts of spent fuel storage are limited to low-level releases of noble gases and iodine. Water needed for water basin storage of spent fuel and transportation accidents are considered; the need to minimize the distance travelled is pointed out. Resource commitments for construction of the storage facilities are analyzed

Partial Defect Verification of Spent Fuel Assemblies by PDET: Principle and Field Testing in Interim Spent Fuel Storage Facility (CLAB) in Sweden

Energy Technology Data Exchange (ETDEWEB)

Ham, Y.S.; Kerr, P.; Sitaraman, S.; Swan, R. [Global Security Directorate, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Rossa, R. [SCK-CEN, Mol (Belgium); Liljenfeldt, H. [SKB in Oskarshamn (Sweden)

2015-07-01

The need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called 'difficult-to-access' areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into 'difficult-to-access' areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reported the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17x17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly burnup levels. (authors)

Radionuclide release from PWR spent fuel specimens with induced cladding defects

International Nuclear Information System (INIS)

Wilson, C.N.; Oversby, V.M.

1984-03-01

Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented. 6 references, 4 figures

Analysis of the Processes in Spent Fuel Pools in Case of Loss of Heat Removal due to Water Leakage

Directory of Open Access Journals (Sweden)

Algirdas Kaliatka

2013-01-01

Full Text Available The safe storage of spent fuel assemblies in the spent fuel pools is very important. These facilities are not covered by leaktight containment; thus, the consequences of overheating and melting of fuel in the spent fuel pools can be very severe. On the other hand, due to low decay heat of fuel assemblies, the processes in pools are slow in comparison with processes in reactor core during LOCA accident. Thus, the accident management measures play a very important role in case of some accidents in spent fuel pools. This paper presents the analysis of possible consequences of fuel overheating due to leakage of water from spent fuel pool. Also, the accident mitigation measure, the late injection of water was evaluated. The analysis was performed for the Ignalina NPP Unit 2 spent fuel pool, using system thermal hydraulic code for severe accident analysis ATHLET-CD. The phenomena, taking place during such accident, are discussed. Also, benchmarking of results of the same accident calculation using ASTEC and RELAP/SCDAPSIM codes is presented here.

Investigation on spent fuel characteristics of reduced-moderation water reactor (RMWR)

International Nuclear Information System (INIS)

Fukaya, Y.; Okubo, T.; Uchikawa, S.

2008-01-01

The spent fuel characteristics of the reduced-moderation water reactor (RMWR) have been investigated using the SWAT and ORIGEN codes. RMWR is an advanced LWR concept for plutonium recycling by using the MOX fuel. In the code calculation, the ORIGEN libraries such as one-group cross-section data prepared for RMWR were necessary. Since there were no open libraries for RMWR, they were produced in this study by using the SWAT code. New libraries based on the heterogeneous core modeling in the axial direction and with the variable actinide cross-section (VXSEC) option were produced and selected as the representative ORIGEN libraries for RMWR. In order to investigate the characteristics of the RMWR spent fuel, the decay heat, the radioactivity and the content of each nuclide were evaluated with ORIGEN using these libraries. In this study, the spent fuel characteristics of other types of reactors, such as PWR, BWR, high burn-up PWR, full-MOX-PWR, full-MOX-BWR and FBR, were also evaluated with ORIGEN. It has been found that about a half of the decay heat of the RMWR spent fuel comes from the actinides nuclides. It is the same with the radioactivity. The decay heat and the radioactivity of the RMWR spent fuel are lower than those of full-MOX-LWRs and FBR, and are the same level as those of the high burn-up PWR. The decay heat and the radioactivity from the fission products (FPs) in the spent fuel mainly depend on the burn-up and the burn-up time rather than the reactor type. Therefore, the decay heat and the radioactivity from FPs in the RMWR spent fuel are smaller, reflecting its relatively long burn-up time resulted from its core characteristics with the high conversion ratio. The radioactivity from the actinides in the spent fuel mainly depends on the 241 Pu content in the initial fuel, and the decay heat mainly depends on 238 Pu and 244 Cm. The contribution of 244 Cm is much smaller in RMWR than in MOX-LWRs because of the difference in the spectrum. In addition, from

Investigation on spent fuel characteristics of reduced-moderation water reactor (RMWR)

Energy Technology Data Exchange (ETDEWEB)

Fukaya, Y. [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency (JAEA), Oarai-machi, Ibaraki-ken 311-1393 (Japan)], E-mail: fukaya.yuji@jaea.go.jp; Okubo, T.; Uchikawa, S. [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency (JAEA), Oarai-machi, Ibaraki-ken 311-1393 (Japan)

2008-07-15

The spent fuel characteristics of the reduced-moderation water reactor (RMWR) have been investigated using the SWAT and ORIGEN codes. RMWR is an advanced LWR concept for plutonium recycling by using the MOX fuel. In the code calculation, the ORIGEN libraries such as one-group cross-section data prepared for RMWR were necessary. Since there were no open libraries for RMWR, they were produced in this study by using the SWAT code. New libraries based on the heterogeneous core modeling in the axial direction and with the variable actinide cross-section (VXSEC) option were produced and selected as the representative ORIGEN libraries for RMWR. In order to investigate the characteristics of the RMWR spent fuel, the decay heat, the radioactivity and the content of each nuclide were evaluated with ORIGEN using these libraries. In this study, the spent fuel characteristics of other types of reactors, such as PWR, BWR, high burn-up PWR, full-MOX-PWR, full-MOX-BWR and FBR, were also evaluated with ORIGEN. It has been found that about a half of the decay heat of the RMWR spent fuel comes from the actinides nuclides. It is the same with the radioactivity. The decay heat and the radioactivity of the RMWR spent fuel are lower than those of full-MOX-LWRs and FBR, and are the same level as those of the high burn-up PWR. The decay heat and the radioactivity from the fission products (FPs) in the spent fuel mainly depend on the burn-up and the burn-up time rather than the reactor type. Therefore, the decay heat and the radioactivity from FPs in the RMWR spent fuel are smaller, reflecting its relatively long burn-up time resulted from its core characteristics with the high conversion ratio. The radioactivity from the actinides in the spent fuel mainly depends on the {sup 241}Pu content in the initial fuel, and the decay heat mainly depends on {sup 238}Pu and {sup 244}Cm. The contribution of {sup 244}Cm is much smaller in RMWR than in MOX-LWRs because of the difference in the spectrum

Studies of the Influence of Water Radiolysis to the Spent Fuel Matrix Dissolution Process

International Nuclear Information System (INIS)

Quinones, J.; Serrano, J.

2001-01-01

The disposal of high level radioactive waste in geological deep repositories relies on the long term stability of spent fuel matrix, which must be assured for thousands of years. One of these factors considered within the studies of performance assessment on spent fuel under final repository conditions is the effect of the radiation on its leaching behaviour. Due to the radiation from spent fuel can modify some properties of both solid phase and leachant and therefore it would alter the chemical behaviour of the near field. Particularizing in the effect of the radiation on the leachant, it will cause generation of radiolytic species that could change the redox potential of the environment and therefore may bring on variations in the leaching process. In this work, we compiled the leaching experiments performed in an irradiation facility (Nayade), in order to emulate γ radiation field of a spent fuel at different cooling times. Initial dose rate used was 0.014 (Gy/s) using source of ''60 Co. The spent fuel chemical analogue utilised was SIMFUEL (natural UO 2 doped with non-radioactive elements simulating fission products) and the leachant selected were saline and granite bentonite waters both under initial anoxic conditions. Preliminary results indicate that radiation produces an increase of the uranium dissolution rate, being the concentrations measured close to those obtained in oxic atmosphere without radiation field. In addition the solubility solid phases from experimental conditions were calculated, for both granite bentonite water and 5 m NaCl media. On the other hand, a tentative approach to model the role of γ radiolysis in these SIMFUEL tests has been carried out as well. (Author)

Simulation of low pressure water hammer

Science.gov (United States)

Himr, D.; Habán, V.

2010-08-01

Numerical solution of water hammer is presented in this paper. The contribution is focused on water hammer in the area of low pressure, which is completely different than high pressure case. Little volume of air and influence of the pipe are assumed in water, which cause sound speed change due to pressure alterations. Computation is compared with experimental measurement.

Characterization of monolith block of spent resin cementation

International Nuclear Information System (INIS)

Prayitno; Endro-Kismolo; Isman MT

1996-01-01

Spent resin immobilization process with cement was done to prevent release of radionuclide in the ultimate storage or disposal. The varied Composition of water/cement ratio in the cementation process were 0.3; 0.4; 0.5 and the various weight of resin waste are 25 g, 37.5 g and 50 gram. The compressive strength of the various water/cement ratio without spent resin was bigger than 0.3. This investigation proved that the compressive strength of Tiga Roda cement was bigger than those of Gresik cement or Nusantara cement. The compressive of the cement block of were the spent resin cementation was influenced by the water/cement ratio and the total spent resin addition. The best condition reached at the water/cement ratio of 0.3 and 25 gram spent resin, was compressive strength of 17.86 N/mm 2 . Leaching rate of the various weight composition of spent resin cementation for 91 days were between 10 -2 - 10 -4 gram.cm -2 .day -1

The influence of near field hydrogen on actinide solubilities and spent fuel leaching

International Nuclear Information System (INIS)

Spahiu, K.; Werme, L.; Eklund, U.B.

2000-01-01

Large amounts of hydrogen are produced as a result of the anoxic corrosion of iron in the proposed container materials for some geologic repositories. Another hydrogen source, less important than the anoxic corrosion of iron, is the radiolysis of water by the spent fuel radiation. Gas phase formation occurs when the pressure of the hydrogen equals at least the hydrostatic pressure, around 5 MPa at 500 meters depth. The effects of 5 MPa hydrogen pressure on spent PWR fuel leaching and on uranium oxide solubility have been studied in carbonated solutions at 70 C. The experiments were performed in a 1 liter autoclave, filled with 950 ml of a solution 10 mM NaCl, 2 mM NaHCO 3 and with hydrogen at a pressure of 5 MPa in the remaining 50 ml free volume. The leaching behavior of 2 g PWR spent fuel powder of the 0.25-0.50 mm fraction, placed in a gold basket was studied during several months by analyzing 10 ml solution samples taken after regular time intervals. A few experiments were performed also with unirradiated U(IV) oxide. In both cases extremely low concentrations of uranium (less than 10 -9 M) were measured in the solution samples. Furthermore the uranium levels in solution remained practically constant during the whole leaching period (more than one year), indicating the absence of any oxidative dissolution of the spent fuel matrix. The same conclusion is confirmed by the constant (within analytical errors) levels of strontium, cesium, molybdenum, iodine and technetium during the whole leaching period. These results have been compared with the ones obtained during the leaching of a spent fuel pin in anoxic conditions, where the uranium and other radionuclides levels are several orders of magnitude higher. The surface of spent fuel or U(IV) oxide is partially oxidized during storage, giving rise to relatively high levels of U(VI) in solution even during leaching in anoxic conditions. No such effect could be observed in the presence of 5 MPa hydrogen, indicating

Burnup credit feasibility for BWR spent fuel shipments

International Nuclear Information System (INIS)

Broadhead, B.L.

1990-01-01

Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent of fuel casks used for transportation and storage. Analyses 1 have shown the feasibility estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This paper summarizes the extension of the previous PWR feasibility assessments to boiling water reactor (BWR) fuel. As with the PWR analysis, the purpose was not verification of burnup credit (see ref. 2 for ongoing work in this area) but a reasonable assessment of the feasibility and potential gains from its use in BWR applications. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. The method includes characterization of a typical pin-cell spectrum, using a one-dimensional (1-D) model of a BWR assembly. The calculated spectrum allows burnup-dependent few-group material constants to be generated. Point depletion methods were then used to obtain the time-varying characteristics of the fuel. These simple methods were validated, where practical, with multidimensional methods. 6 refs., 1 tab

Flashing of high-pressure saturated water into the pool water

International Nuclear Information System (INIS)

Takamasa, Tomoji; Kondo, Koichi; Aya, Izuo.

1997-01-01

This paper presents an experimental study on a saturated high-pressure water discharging into a water pool. The purpose of the experiment is to clarify the phenomena that occur by a blow-down of the water from the pressure vessel into the water-filled containment in the case of a wall-crack accident or a LOCA in a passive safety reactor. The results show that a flashing oscillation (FO) occurs when the water discharges into the pool, under specified experimental conditions. The range of the flashing location oscillates between a point very close to and some distance away from the vent hole. The pressures in the vent tube and water pool constantly fluctuate due to the flashing oscillation. The pressure oscillation and alternating flashing location might be caused by the balancing action between the supply of saturated water, flashing at the control volume and steam condensation on the steam-water interface. The frequencies of FO, or frequencies of pressure oscillation and alternating flashing location, increased as water subcooling increased, and as discharging pressure and vent hole diameter decreased. A linear analysis was conducted using a spherical flashing bubble model in which the motion of bubble is controlled by steam condensation. The effects of these parameters on the period of FO in the experiments can be predicted well by the analysis. (author)

Analyses of the transportation of spent research reactor fuel in the United States

International Nuclear Information System (INIS)

Cashwell, J.W.; Neuhauser, K.S.

1989-01-01

We analyzed the impacts of transportation of research reactor spent fuel from US and foreign reactors for the US Department of Energy's (DOE) Office of Defense Programs. Two separate shipment programs were analyzed. The shipment of research reactor spent fuel from Taiwan to the US (Fuel Movement Program), and the return of research reactor spent fuels of US origin from foreign and domestic reactors (Research Reactor Fuel Return Program). To perform these analyses, a comprehensive methodology for analyzing the probabilities and consequences of transportation in coastal waters and port facilities, handling at the port, and shipment by truck to reprocessing facilities was developed. The Taiwanese fuel consists of low-burnup aluminum-clad metallic uranium research reactor spent fuel; the other fuels are primarily aluminum-clad oxide fuels. The Fuel Movement Program is ongoing, while the Fuel Return Program addresses future shipments over a ten-year period. The operational aspects of the Taiwanese shipments have been uniform, but several possible shipping configurations are possible for the Fuel Return Program shipments. The risks of transporting spent nuclear fuel and other radioactive materials by all modes have been analyzed extensively. Comprehensive assessments, which bound the impacts of spent fuel transport, demonstrate that when shipments are made in compliance with applicable regulations, the risks for all such transport are low. For comparison with previously licensed transport activities and to provide continuity with earlier analyses, the results for shipment of 150-day-old commercial pressurized water reactor (PWR) spent fuel are presented as part of this study

Spent fuel receipt and lag storage facility for the spent fuel handling and packaging program

International Nuclear Information System (INIS)

Black, J.E.; King, F.D.

1979-01-01

Savannah River Laboratory (SRL) is participating in the Spent Fuel Handling and Packaging Program for retrievable, near-surface storage of spent light water reactor (LWR) fuel. One of SRL's responsibilities is to provide a technical description of the wet fuel receipt and lag storage part of the Spent Fuel Handling and Packaging (SFHP) facility. This document is the required technical description

Yugoslav spent nuclear fuel management program and international perspectives

International Nuclear Information System (INIS)

Pesic, M.; Subotic, K.; Sotic, O.; Plecas, I.; Ljubenov, V.; Peric, A.; Milosevic, M.

2002-01-01

Spent nuclear fuel stored in the Vinca Institute of Nuclear Sciences, Yugoslavia, consists of about 2.5 tons of metal uranium (initial enrichment 2%) and about 20 kg uranium dioxide (dispersed in aluminum matrix, initial fuel uranium enrichment 80%). This spent nuclear fuel is generated in operation of the RA heavy water research reactor during 1959-1984 period. Both types of fuel are of ex-USSR origin, have the same shape and dimensions and approximately the same initial mass of 235 nuclide. They are known as the TVR-S type of fuel elements. The total of 8030 spent fuel elements are stored at the RA research reactor premises, almost all in the spent fuel pool filled by ordinary water. The last used 480 high-enriched uranium spent fuel elements are kept in the drained RA reactor core since 1984. Fuel layer of both enrichments is covered with thin aluminium cladding. Due to non-suitable chemical parameters of water in the spent fuel storage pool, the corrosion processes penetrated aluminium cladding and aluminium walls od storage containers during storage period long from 20 to 40 years. Activity of fission products ( 137 Cs) is detected in water samples during water inspection in 1996 and experts of the lAEA Russia and USA were invited to help. By end of 2001, some remediation of the water transparency of the storage pool and inspections of water samples taken from the storage containers with the spent fuel elements were carried out by the Vinca Institute staff and with the help of experts from the Russia and the IAEA. Following new initiatives on international perspective on spent fuel management, a proposal was set by the IAEA, and was supported by the governments of the USA and the Russian Federation to ship the spent fuel elements of the RA research reactor to Mayak spent fuel processing plant in Russia. This paper describes current status of the reactor RA spent fuel elements, initiative for new Yugoslav spent fuel management program speculates on some of the

Pressurized-water reactors

International Nuclear Information System (INIS)

Bush, S.H.

1983-03-01

An overview of the pressurized-water reactor (PWR) pressure boundary problems is presented. Specifically exempted will be discussions of problems with pumps, valves and steam generators on the basis that they will be covered in other papers. Pressure boundary reliability is examined in the context of real or perceived problems occurring over the past 5 to 6 years since the last IAEA Reliability Symposium. Issues explicitly covered will include the status of the pressurized thermal-shock problem, reliability of inservice inspections with emphasis on examination of the region immediately under the reactor pressure vessel (RPV) cladding, history of piping failures with emphasis on failure modes and mechanisms. Since nondestructive examination is the topic of one session, discussion will be limited to results rather than techniques

Current perceptions of spent nuclear fuel behavior in water pool storage

International Nuclear Information System (INIS)

Johnson, A.B. Jr.

1977-06-01

A survey was conducted of a cross section of U.S. and Canadian fuel storage pool operators to define the spent fuel behavior and to establish the range of pool storage environments. There is no evidence for significant corrosion degradation. Fuel handling causes only minimal damage. Most fuel bundles with defects generally are stored without special procedures. Successful fuel storage up to 18 years with benign water chemistry has been demonstrated. 2 tables

Effects of Burnable Absorbers on PWR Spent Nuclear Fuel

International Nuclear Information System (INIS)

O'Leary, P.M.; Pitts, M.L.

2000-01-01

Burnup credit is an ongoing issue in designing and licensing transportation and storage casks for spent nuclear fuel (SNF). To address this issue, in July 1999, the U.S. Nuclear Regulatory Commission (NRC), Spent Fuel Project Office, issued Interim Staff Guidance-8 (ISG-8), Revision 1 allowing limited burnup credit for pressurized water reactor (PWR) spent nuclear fuel (SNF) to be used in transport and storage casks. However, one of the key limitations for a licensing basis analysis as stipulated in ISG-8, Revision 1 is that ''burnup credit is restricted to intact fuel assemblies that have not used burnable absorbers''. Because many PWR fuel designs have incorporated burnable-absorber rods for more than twenty years, this restriction places an unnecessary burden on the commercial nuclear power industry. This paper summarizes the effects of in-reactor irradiation on the isotopic inventory of PWR fuels containing different types of integral burnable absorbers (BAs). The work presented is illustrative and intended to represent typical magnitudes of the reactivity effects from depleting PWR fuel with different types of burnable absorbers

The security management of spent filter cartridge in Qinshan phase 3 (heavy water reactor) nuclear power plant

International Nuclear Information System (INIS)

Xue Dahai

2005-01-01

Qinshan phase 3 nuclear power plant is the first CANDU plant that China fetched in from Canada, and both two units operate under well condition up to now. The radioactive wastes produced during the unit operation mainly include technical waste, spent resin, and spent filter cartridge. The spent filter cartridge is one important part both in the volume and radioactivity of the radioactive waste, and it is the important content of radioactive waste management. Different from PWR, part of high radioactive spent filter in CANDU unit comes from heavy water system such as moderator system. It has to be dried through blowing before replaced from the system. But this working procedure result the filtrate dreg become flexible, and it can bring on the risk of internal or external exposure. It is very important to pay high attention to control the contamination spread during spent filter inside transfer. (authors)

Development of a strategy for the management of PBMR spent fuel in South Africa - HTR2008-58047

International Nuclear Information System (INIS)

Smith, S. W.; Bredell, P. J.; Meyer, W. C. M. H.

2008-01-01

South Africa is planning to expand its nuclear power generating capacity by deploying a number of pressurized-water reactors and pebble-bed modular reactors. It can be expected that this program will impact on the current and planned spent fuel and radioactive waste management systems in South Africa. This paper proposes an approach to develop a strategy for the management of PBMR spent fuel that would contribute to the optimization of the overall national radwaste management system. The approach is expected to provide a conceptual spent fuel management strategy and will also highlight areas that need to be further developed, thus providing guidance for basic technology development. (authors)

Uranium-236 in light water reactor spent fuel recycled to an enriching plant

International Nuclear Information System (INIS)

de la Garza, A.

1977-01-01

The introduction of 236 U to an enriching plant by recycling spent fuel uranium results in enriched products containing 236 U, a parasitic neutron absorber in reactor fuel. Convenient approximate methodology determines 235 236 U, and total uranium flowsheets with associated separative work requirements in enriching plant operations for use by investigators of the light water reactor fuel cycle not having recourse to specialized multicomponent cascade technology. Application of the methodology has been made to compensation of an enriching plant product for 236 U content and to the value at an enriching plant of spent fuel uranium. The approximate methodology was also confirmed with more exact calculations and with some experience with 236 U in an enriching plant

Deep geological disposal system development; mechanical structural stability analysis of spent nuclear fuel disposal canister under the internal/external pressure variation

Energy Technology Data Exchange (ETDEWEB)

Kwen, Y. J.; Kang, S. W.; Ha, Z. Y. [Hongik University, Seoul (Korea)

2001-04-01

This work constitutes a summary of the research and development work made for the design and dimensioning of the canister for nuclear fuel disposal. Since the spent nuclear fuel disposal emits high temperature heats and much radiation, its careful treatment is required. For that, a long term(usually 10,000 years) safe repository for spent fuel disposal should be securred. Usually this repository is expected to locate at a depth of 500m underground. The canister construction type introduced here is a solid structure with a cast iron insert and a corrosion resistant overpack, which is designed for spent nuclear fuel disposal in a deep repository in the crystalline bedrock, which entails an evenly distributed load of hydrostatic pressure from undergroundwater and high pressure from swelling of bentonite buffer. Hence, the canister must be designed to withstand these high pressure loads. Many design variables may affect the structural strength of the canister. In this study, among those variables array type of inner baskets and thicknesses of outer shell and lid and bottom are tried to be determined through the mechanical linear structural analysis, thicknesses of outer shell is determined through the nonlinear structural analysis, and the bentonite buffer analysis for the rock movement is conducted through the of nonlinear structural analysis Also the thermal stress effect is computed for the cast iron insert. The canister types studied here are one for PWR fuel and another for CANDU fuel. 23 refs., 60 figs., 23 tabs. (Author)

High pressure experimental water loop

International Nuclear Information System (INIS)

Grenon, M.

1958-01-01

A high pressure experimental water loop has been made for studying the detection and evolution of cladding failure in a pressurized reactor. The loop has been designed for a maximum temperature of 360 deg. C, a maximum of 160 kg/cm 2 and flow rates up to 5 m 3 /h. The entire loop consists of several parts: a main circuit with a canned rotor circulation pump, steam pressurizer, heating tubes, two hydro-cyclones (one de-gasser and one decanter) and one tubular heat exchanger; a continuous purification loop, connected in parallel, comprising pressure reducing valves and resin pots which also allow studies of the stability of resins under pressure, temperature and radiation; following the gas separator is a gas loop for studying the recombination of the radiolytic gases in the steam phase. The preceding circuits, as well as others, return to a low pressure storage circuit. The cold water of the low pressure storage flask is continuously reintroduced into the high pressure main circuit by means of a return pump at a maximum head of 160 kg /cm 2 , and adjusted to the pressurizer level. This loop is also a testing bench for the tight high pressure apparatus. The circulating pump and the connecting flanges (Oak Ridge type) are water-tight. The feed pump and the pressure reducing valves are not; the un-tight ones have a system of leak recovery. To permanently check the tightness the circuit has been fitted with a leak detection system (similar to the HRT one). (author) [fr

Effects of actinide compositional variability in the US spent fuel inventory on partitioning-transmutation systems

International Nuclear Information System (INIS)

Ludwig, S.B.; Michaels, G.E.; Hanson, B.D.

1992-01-01

Partitioning and transmutation (P-T) is an advanced waste management concept by which certain undesirable nuclides in spent fuel are first isolated (partitioned) and later destroyed (transmuted) in a nuclear reactor or other transmutation device. There are wide variabilities in the nuclide composition of spent fuel. This implies that there will also be wide variabilities in the transmutation device feed. As a waste management system, P-T must be able to accept (all) spent fuel. Variability of nuclide composition (i.e., the feed material for transmutation devices) may be important because virtually all transmutation systems propose to configure transuranic (TRU) nuclides recovered from discharged lightwater reactor (LWR) spent fuel in critical or near-critical cores. To date, all transmutation system core analyses assume invariant nuclide concentrations for startup and recycle cores. Using the US Department of Energy's (DOE's) Characteristics Data Base (CDB) and the ORIGEN2 computer code, the current and projected spent fuel discharges until the year 2016 have been categorized according to combinations of fuel burnup, initial enrichment, fuel age (cooling time) and reactor type (boiling-water or pressurized-water reactors). The variability of the infinite multiplication factor (k ∞ ) is calculated for both fast (ALMR) and thermal (accelerator-based) transmuter systems

Purification of polluted water with spent mushroom (Agaricus bisporus) substrate: from agricultural waste to biosorbent of phenanthrene, Cd and Pb.

Science.gov (United States)

García-Delgado, C; Alonso-Izquierdo, M; González-Izquierdo, M; Yunta, F; Eymar, E

2017-07-01

The present research was aimed to (i) report the recycling of spent A. bisporus substrate (SAS) to remove heavy metals (Cd and Pb) and phenanthrene (Phe) from polluted water and (ii) assess the possibility to use the treated water for irrigation. Batch experiments were carried out to assess, firstly, the effect of interaction time between pollutants with SAS and, secondly, the pH of the polluted water. Then a biofilter was designed by using pressurized glass columns. Chemical parameters such as pH, electrical conductivity and content of Pb, Cd, Phe, nutrients (NPK) and Cl - were determined. Equilibrium for contaminants was quickly reached (1-2 h). The pH of the polluted water was the key factor for pollutants' adsorption. The polluted water's pH was increased after biofilter interaction. Phe was not detected in any fraction. Pb and Cd sorption rates were higher than 99%. The pollutant concentrations were within the permitted range to be used for agriculture purposes. Purified water showed significant concentrations of NPK, indicating its potential use as fertilizer. The SAS shows potential to be used as Phe, Pb and Cd biosorbent and the resulting treated water can be used for irrigation according to pollutant contents and agronomical evaluation.

Nondestructive assay of spent fuel rods from a Light Water Breeder Reactor (LWBR Development Program)

International Nuclear Information System (INIS)

Tessler, G.; Beaudoin, B.R.; Beggs, W.J.; Freeman, L.B.; Schick, W.C. Jr.

1987-09-01

A gauge, called the Production Irradiated Fuel Assay Gauge (PIFAG), has been developed and utilized to measure, nondestructively, the fissile fuel content of spent fuel rods from the Light Water Breeder Reactor (LWBR) core. The PIFAG was in operation from November 1983 to May 1987. During this period, assay data were obtained for two irradiated test rods used for initial qualification of the gauge and 524 spent LWBR core rods. A review of PIFAG operations is given, including hot cell operations, calibration, assay operations, and methods used to monitor the data quality and verify the precision and accuracy of the data. The analytical model used to determine the core rod fissile fuel content from the data and the results for the 524 LWBR spent fuel rods are given

Flow evaluation of the leaching hazardous materials from spent nickel-cadmium batteries discarded in different water surroundings.

Science.gov (United States)

Guo, Xingmei; Song, Yan; Nan, Junmin

2018-02-01

The leaching characteristics of hazardous materials from Ni-Cd batteries immersed in four typical water samples, i.e., water with NaCl, river water, tap water, and deionized water, were investigated to evaluate the potential environmental harm of spent Ni-Cd batteries in the water surroundings. It is shown that four water surroundings all could leach hazardous materials from the Ni-Cd batteries. The water with NaCl concentration of 66.7 mg L -1 had the highest leaching ability, the hazardous materials were leached after only approximately 50 days (average time, with a standard deviation of 4.1), while less than 100 days were needed in the others. An electrochemical corrosion is considered to be the main leaching mechanism leading to battery breakage, while the dissolution-deposition process and the powder route result in the leakage and transference of nickel and cadmium materials from the electrodes. The anions, i.e., SO 4 2- and Cl - , and dissolved oxygen in water were demonstrated to be the vital factors that influence the leaching processes. Thus, it is proposed that spent Ni-Cd batteries must be treated properly to avoid potential danger to the environment.

Analyses of the transportation of spent research reactor fuel in the United States

International Nuclear Information System (INIS)

Cashwell, J.W.; Neuhauser, K.S.

1989-01-01

The Transportation Technology Center at Sandia National Laboratories has analyzed the impacts of transportation of research reactor spent fuel from US and foreign reactors for the US Department of Energy (DOE) Office of Defense Programs. This effort represents the first comprehensive analytical evaluation of the risks of transporting high-, medium-, and low-enriched uranium spent research reactor fuel by both sea and land. Two separate shipment programs have been analyzed: the shipment of research reactor spent fuel from Taiwan to the US (Fuel Movement Program), and the return of research reactor spent fuels of US origin from foreign and domestic reactors (Research Reactor Fuel Return Program). In order to perform these analyses, a comprehensive methodology for analyzing the probabilities and consequences of transportation in coastal waters and port facilities, handling at the port, and shipment by truck to reprocessing facilities was developed. The Taiwanese fuel consists of low-burnup aluminum-clad metallic uranium research reactor spent fuel; the other fuels are primarily aluminum-clad oxide fuels. The Fuel Movement Program is ongoing, while the Fuel Return Program addresses future shipments over a ten-year period. The operational aspects of the Taiwanese shipments have been uniform, but several possible shipping configurations are possible for the Fuel Return Program shipments. Comprehensive assessments, which bound the impacts of spent fuel transport, demonstrate that when shipments are made in compliance with applicable regulations, the risks for all such transport are low. For comparison with previously licensed transport activities and to provide continuity with earlier analyses, the results for shipment of 150-day-old commercial pressurized water reactor (PWR) spent fuel are presented as part of this study

Investigation of the condition of spent-fuel pool components

International Nuclear Information System (INIS)

Kustas, F.M.; Bates, S.O.; Opitz, B.E.; Johnson, A.B. Jr.; Perez, J.M. Jr.; Farnsworth, R.K.

1981-09-01

It is currently projected that spent nuclear fuel, which is discharged from the reactor and then stored in water pools, may remain in those pools for several decades. Other studies have addressed the expected integrity of the spent fuel during extended water storage; this study assesses the integrity of metallic spent fuel pool components. Results from metallurgical examinations of specimens taken from stainless steel and aluminum components exposed in spent fuel pools are presented. Licensee Event Reports (LERs) relating to problems with spent fuel components were assessed and are summarized to define the types of operational problems that have occurred. The major conclusions of this study are: aluminum and stainless steel spent fuel pool components have a good history of performance in both deionized and borated water pools. Although some operational problems involving pool components have occurred, these problems have had minimal impacts

Investigation of the condition of spent-fuel pool components

Energy Technology Data Exchange (ETDEWEB)

Kustas, F.M.; Bates, S.O.; Opitz, B.E.; Johnson, A.B. Jr.; Perez, J.M. Jr.; Farnsworth, R.K.

1981-09-01

It is currently projected that spent nuclear fuel, which is discharged from the reactor and then stored in water pools, may remain in those pools for several decades. Other studies have addressed the expected integrity of the spent fuel during extended water storage; this study assesses the integrity of metallic spent fuel pool components. Results from metallurgical examinations of specimens taken from stainless steel and aluminum components exposed in spent fuel pools are presented. Licensee Event Reports (LERs) relating to problems with spent fuel components were assessed and are summarized to define the types of operational problems that have occurred. The major conclusions of this study are: aluminum and stainless steel spent fuel pool components have a good history of performance in both deionized and borated water pools. Although some operational problems involving pool components have occurred, these problems have had minimal impacts.

Natural convection heat transfer within horizontal spent nuclear fuel assemblies

International Nuclear Information System (INIS)

Canaan, R.E.

1995-12-01

Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array

Alternatives for water basin spent fuel storage: executive summary and comparative evaluation

International Nuclear Information System (INIS)

Viebrock, J.M.

1979-09-01

A five part report identifies and evaluates alternatives to conventional methods for water basin storage of irradiated light water reactor fuel assemblies (spent fuel). A recommendation is made for development or further evaluation of one attractive alternative: Proceed to develop fuel disassembly with subsequent high density storage of fuel pins (pin storage). The storage alternatives were evaluated for emplacement at reactor, in existing away-from-reactor storage facilities and in new away-from-reactor facilities. In the course of the study, the work effort necessarily extended beyond the pool wall in scope to properly assess the affects of storage alternatives on AFT systems

Analysis of the loss of pool cooling accident in a PWR spent fuel pool with MAAP5

International Nuclear Information System (INIS)

Wu, Xiaoli; Li, Wei; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

2014-01-01

Highlights: • A PWR spent fuel pool was modeled by using MAAP5. • Loss of pool cooling severe accident scenarios were studied. • Loss of pool cooling accidents with two mitigation measures were analyzed. - Abstract: The Fukushima Daiichi nuclear accident shows that it is necessary to study potential severe accidents and corresponding mitigation measures for the spent fuel pool (SFP) of a nuclear power plant (NPP). This paper presents the analysis of loss of pool cooling accident scenarios and the discussion of mitigation measures for the SFP at a pressurized water reactor (PWR) NPP with the MAAP5 code. Analysis of uncompensated loss of water due to the loss of pool cooling with different initial pool water levels of 12.2 m (designated as a reference case) and 10.7 m have been performed based on a MAAP5 input model. Scenarios of the accident such as overheating of uncovered fuel assemblies, oxidation of claddings and hydrogen generation, loss of intactness of fuel rod claddings, and release of radioactive fission products were predicted with the assumption that mitigation measures were unavailable. The results covered a broad spectrum of severe accident evaluations in the SFP. Furthermore, as important mitigation measures, the effects of recovering the SFP cooling system and makeup water in SFP on the accident progressions have also been investigated respectively based on the events of pool water boiling and spent fuels uncovery. Based upon the reference case, three cases with the recovery of SFP cooling system and three other cases with makeup water in SFP have been studied. The results showed that, severe accident might happen if SFP cooling system was not restored timely before the spent fuels started to become uncovered; spent fuels could be completely submerged and severe accident might be avoided if SFP makeup water system provided water with a mass flow rate larger than the average evaporation rate defined as the division of pool water mass above the

Automatic spent fuel ID number reader (I)

International Nuclear Information System (INIS)

Tanabe, S.; Kawamoto, H.; Fujimaki, K.; Kobe, A.

1991-01-01

An effective and efficient technique has been developed for facilitating identification works of LWR spent fuel stored in large scale spent fuel storage pools of such as processing plants. Experience shows that there are often difficulties in the implementation of operator's nuclear material accountancy and control works as well as safeguards inspections conducted on spent fuel assemblies stored in deep water pool. This paper reports that the technique is realized as an automatic spent fuel ID number reader system installed on fuel handling machine. The ID number reader system consists of an optical sub-system and an image processing sub-system. Thousands of spent fuel assemblies stored in under water open racks in each storage pool could be identified within relatively short time (e.g. within several hours) by using this combination. Various performance tests were carried out on image processing sub-system in 1990 using TV images obtained from different types of spent fuel assemblies stored in various storage pools of PWR and BWR power stations

Water-Pressure Distribution on Seaplane Float

Science.gov (United States)

Thompson, F L

1929-01-01

The investigation presented in this report was conducted for the purpose of determining the distribution and magnitude of water pressures likely to be experienced on seaplane hulls in service. It consisted of the development and construction of apparatus for recording water pressures lasting one one-hundredth second or longer and of flight tests to determine the water pressures on a UO-1 seaplane float under various conditions of taxiing, taking off, and landing. The apparatus developed was found to operate with satisfactory accuracy and is suitable for flight tests on other seaplanes. The tests on the UO-1 showed that maximum pressures of about 6.5 pounds per square inch occur at the step for the full width of the float bottom. Proceeding forward from the step the maximum pressures decrease in magnitude uniformly toward the bow, and the region of highest pressures narrows toward the keel. Immediately abaft the step the maximum pressures are very small, but increase in magnitude toward the stern and there once reached a value of about 5 pounds per square inch. (author)

LWR spent-fuel radiochemical measurements and comparison with ORIGEN2 predictions

International Nuclear Information System (INIS)

Blahnik, D.E.; Jenquin, U.P.; Guenther, R.J.

1988-01-01

The Materials Characterization Center (MCC) at Pacific Northwest Laboratory is responsible for providing characterized spent fuel, designated approved testing material (ATMs) for subsequent use in the investigation of nuclear waste disposal forms by the US Department of Energy geologic repository project. The ATMs are selected to assure that test material is available that has a representative range of characteristics important to spent-fuel behavior in a geologic repository. Burnup and fission gas release were the primary criteria for selecting the ATMs. The five spent-fuel ATMs (ATM-101, -103, -104, -105, and -106) currently being characterized by the MCC have rod average burnups ranging from 20 to 43 MWd/kg M, fission gas releases ranging from 0.2 to 11.2%, and are from both boiling water reactors and pressurized water reactors. Radiochemical analyses of the fuel included measurements of 148 Nd (for fuel burnup), the isotopes of uranium and plutonium, and nuclides of importance to repository performance. Cladding samples were analyzed for 14 C. The measured values of selected nuclides were compared with values obtained from calculations with the ORIGEN2 code that was used to predict isotopic quantities for all of the ATMS. Ratios of the ORIGEN2 calculated values to the measured values for ATM-103 and ATM-106 fuel are given

Spent fuel transportation in the United States: commercial spent fuel shipments through December 1984

International Nuclear Information System (INIS)

1986-04-01

This report has been prepared to provide updated transportation information on light water reactor (LWR) spent fuel in the United States. Historical data are presented on the quantities of spent fuel shipped from individual reactors on an annual basis and their shipping destinations. Specifically, a tabulation is provided for each present-fuel shipment that lists utility and plant of origin, destination and number of spent-fuel assemblies shipped. For all annual shipping campaigns between 1980 and 1984, the actual numbers of spent-fuel shipments are defined. The shipments are tabulated by year, and the mode of shipment and the casks utilized in shipment are included. The data consist of the current spent-fuel inventories at each of the operating reactors as of December 31, 1984. This report presents historical data on all commercial spent-fuel transportation shipments have occurred in the United States through December 31, 1984

High Pressure Industrial Water Facility

Science.gov (United States)

1992-01-01

In conjunction with Space Shuttle Main Engine testing at Stennis, the Nordberg Water Pumps at the High Pressure Industrial Water Facility provide water for cooling the flame deflectors at the test stands during test firings.

Contribution of water vapor pressure to pressurization of plutonium dioxide storage containers

Science.gov (United States)

Veirs, D. Kirk; Morris, John S.; Spearing, Dane R.

2000-07-01

Pressurization of long-term storage containers filled with materials meeting the US DOE storage standard is of concern.1,2 For example, temperatures within storage containers packaged according to the standard and contained in 9975 shipping packages that are stored in full view of the sun can reach internal temperatures of 250 °C.3 Twenty five grams of water (0.5 wt.%) at 250 °C in the storage container with no other material present would result in a pressure of 412 psia, which is limited by the amount of water. The pressure due to the water can be substantially reduced due to interactions with the stored material. Studies of the adsorption of water by PuO2 and surface interactions of water with PuO2 show that adsorption of 0.5 wt.% of water is feasible under many conditions and probable under high humidity conditions.4,5,6 However, no data are available on the vapor pressure of water over plutonium dioxide containing materials that have been exposed to water.

Spent nuclear fuel in Bulgaria

International Nuclear Information System (INIS)

Peev, P.; Kalimanov, N.

1999-01-01

The development of the nuclear energy sector in Bulgaria is characterized by two major stages. The first stage consisted of providing a scientific basis for the programme for development of the nuclear energy sector in the country and was completed with the construction of an experimental water-water reactor. At present, spent nuclear fuel from this reactor is placed in a water filled storage facility and will be transported back to Russia. The second stage consisted of the construction of the 6 NPP units at the Kozloduy site. The spent nuclear fuel from the six units is stored in at reactor pools and in an additional on-site storage facility which is nearly full. In order to engage the government of the country with the on-site storage problems, the new management of the National Electric Company elaborated a policy on nuclear fuel cycle and radioactive waste management. The underlying policy is de facto the selection of the 'deferred decision' option for its spent fuel management. (author)

PWR [pressurized water reactor] pressurizer transient response: Final report

International Nuclear Information System (INIS)

Murphy, S.I.

1987-08-01

To predict PWR pressurizer transients, Ahl proposed a three region model with a universal coefficient to represent condensation on the water surface. Specifically, this work checks the need for three regions and the modeling of the interfacial condensation coefficient. A computer model has been formulated using the basic mass and energy conservation laws. A two region vapor and liquid model was first used to predict transients run on a one-eleventh scale Freon pressurizer. These predictions verified the need for a second liquid region. As a result, a three region model was developed and used to predict full-scale pressurizer transients at TMI-2, Shippingport, and Stade. Full-scale pressurizer predictions verified the three region model and pointed out the shortcomings of Ahl's universal condensation coefficient. In addition, experiments were run using water at low pressure to study interface condensation. These experiments showed interface condensation to be significant only when spray flow is turned on; this result was incorporated in the final three region model

Control of degradation of spent LWR [light-water reactor] fuel during dry storage in an inert atmosphere

International Nuclear Information System (INIS)

Cunningham, M.E.; Simonen, E.P.; Allemann, R.T.; Levy, I.S.; Hazelton, R.F.

1987-10-01

Dry storage of Zircaloy-clad spent fuel in inert gas (referred to as inerted dry storage or IDS) is being developed as an alternative to water pool storage of spent fuel. The objectives of the activities described in this report are to identify potential Zircaloy degradation mechanisms and evaluate their applicability to cladding breach during IDS, develop models of the dominant Zircaloy degradation mechanisms, and recommend cladding temperature limits during IDS to control Zircaloy degradation. The principal potential Zircaloy cladding breach mechanisms during IDS have been identified as creep rupture, stress corrosion cracking (SCC), and delayed hydride cracking (DHC). Creep rupture is concluded to be the primary cladding breach mechanism during IDS. Deformation and fracture maps based on creep rupture were developed for Zircaloy. These maps were then used as the basis for developing spent fuel cladding temperature limits that would prevent cladding breach during a 40-year IDS period. The probability of cladding breach for spent fuel stored at the temperature limit is less than 0.5% per spent fuel rod. 52 refs., 7 figs., 1 tab

In-Field Performance Testing of the Fork Detector for Quantitative Spent Fuel Verification

International Nuclear Information System (INIS)

Gauld, Ian C.; Hu, Jianwei; De Baere, P.; Tobin, Stephen

2015-01-01

Expanding spent fuel dry storage activities worldwide are increasing demands on safeguards authorities that perform inspections. The European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) require measurements to verify declarations when spent fuel is transferred to difficult-to-access locations, such as dry storage casks and the repositories planned in Finland and Sweden. EURATOM makes routine use of the Fork detector to obtain gross gamma and total neutron measurements during spent fuel inspections. Data analysis is performed by modules in the integrated Review and Analysis Program (iRAP) software, developed jointly by EURATOM and the IAEA. Under the framework of the US Department of Energy-EURATOM cooperation agreement, a module for automated Fork detector data analysis has been developed by Oak Ridge National Laboratory (ORNL) using the ORIGEN code from the SCALE code system and implemented in iRAP. EURATOM and ORNL recently performed measurements on 30 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel (Clab), operated by the Swedish Nuclear Fuel and Waste Management Company (SKB). The measured assemblies represent a broad range of fuel characteristics. Neutron count rates for 15 measured pressurized water reactor assemblies are predicted with an average relative standard deviation of 4.6%, and gamma signals are predicted on average within 2.6% of the measurement. The 15 measured boiling water reactor assemblies exhibit slightly larger deviations of 5.2% for the gamma signals and 5.7% for the neutron count rates, compared to measurements. These findings suggest that with improved analysis of the measurement data, existing instruments can provide increased verification of operator declarations of the spent fuel and thereby also provide greater ability to confirm integrity of an assembly. These results support the application of the Fork detector as a fully quantitative spent fuel

In-Field Performance Testing of the Fork Detector for Quantitative Spent Fuel Verification

Energy Technology Data Exchange (ETDEWEB)

Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); De Baere, P. [European Commission (Luxembourg). DG Energy, Directorate Nuclear Safeguards; Vaccaro, S. [European Commission (Luxembourg). DG Energy, Directorate Nuclear Safeguards; Schwalbach, P. [European Commission (Luxembourg). DG Energy, Directorate Nuclear Safeguards; Liljenfeldt, Henrik [Swedish Nuclear Fuel and Waste Management Company (Sweden); Tobin, Stephen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-01-01

Expanding spent fuel dry storage activities worldwide are increasing demands on safeguards authorities that perform inspections. The European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) require measurements to verify declarations when spent fuel is transferred to difficult-to-access locations, such as dry storage casks and the repositories planned in Finland and Sweden. EURATOM makes routine use of the Fork detector to obtain gross gamma and total neutron measurements during spent fuel inspections. Data analysis is performed by modules in the integrated Review and Analysis Program (iRAP) software, developed jointly by EURATOM and the IAEA. Under the framework of the US Department of Energy–EURATOM cooperation agreement, a module for automated Fork detector data analysis has been developed by Oak Ridge National Laboratory (ORNL) using the ORIGEN code from the SCALE code system and implemented in iRAP. EURATOM and ORNL recently performed measurements on 30 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel (Clab), operated by the Swedish Nuclear Fuel and Waste Management Company (SKB). The measured assemblies represent a broad range of fuel characteristics. Neutron count rates for 15 measured pressurized water reactor assemblies are predicted with an average relative standard deviation of 4.6%, and gamma signals are predicted on average within 2.6% of the measurement. The 15 measured boiling water reactor assemblies exhibit slightly larger deviations of 5.2% for the gamma signals and 5.7% for the neutron count rates, compared to measurements. These findings suggest that with improved analysis of the measurement data, existing instruments can provide increased verification of operator declarations of the spent fuel and thereby also provide greater ability to confirm integrity of an assembly. These results support the application of the Fork detector as a fully quantitative spent fuel

Some implications of batch average burnup calculations on predicted spent fuel compositions

International Nuclear Information System (INIS)

Alexander, C.W.; Croff, A.G.

1984-01-01

The accuracy of using batch-averaged burnups to determine spent fuel characteristics (such as isotopic composition, activity, etc.) was examined for a typical pressurized-water reactor (PWR) fuel discharge batch by comparing characteristics computed by (a) performing a single depletion calculation using the average burnup of the spent fuel and (b) performing separate depletion calculations based on the relative amounts of spent fuel in each of twelve burnup ranges and summing the results. The computations were done using ORIGEN 2. Procedure (b) showed a significant shift toward a greater quantity of the heavier transuranics, which derive from multiple neutron captures, and a corresponding decrease in the amounts of lower transuranics. Those characteristics which derive primarily from fission products, such as total radioactivity and total thermal power, are essentially identical for the two procedures. Those characteristics that derive primarily from the heavier transuranics, such as spontaneous fission neutrons, are underestimated by procedure (a)

Characterization of spent fuel approved testing material: ATM-106

International Nuclear Information System (INIS)

Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thornhill, C.K.

1988-10-01

The characterization data obtained to date are described for Approved Testing Material (ATM)-106 spent fuel from Assembly BT03 of pressurized-water reactor Calvert Cliffs No. 1. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well- characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCWRM) program. ATM-106 consists of 20 full-length irradiated fuel rods with rod-average burnups of about 3700 GJ/kgM (43 MWd/kgM) and expected fission gas release of /approximately/10%. Characterization data include (1) as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) calculated nuclide inventories and radioactivities in the fuel and cladding; and (6) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel rod are being conducted and will be included in planned revisions of this report. 12 refs., 110 figs., 81 tabs

Measurement techniques in dry-powdered processing of spent nuclear fuels

International Nuclear Information System (INIS)

Bowers, D. L.; Hong, J.-S.; Kim, H.-D.; Persiani, P. J.; Wolf, S. F.

1999-01-01

High-performance liquid chromatography (HPLC) with inductively coupled plasma mass spectrometry (ICPMS) detection, α-spectrometry (α-S), and γ-spectrometry (γ-S) were used for the determination of nuclide content in five samples excised from a high-burnup fuel rod taken from a pressurized water reactor (PWR). The samples were prepared for analysis by dissolution of dry-powdered samples. The measurement techniques required no separation of the plutonium, uranium, and fission products. The sample preparation and analysis techniques showed promise for in-line analysis of highly-irradiated spent fuels in a dry-powdered process. The analytical results allowed the determination of fuel burnup based on 148 Nd, Pu, and U content. A goal of this effort is to develop the HPLC-ICPMS method for direct fissile material accountancy in the dry-powdered processing of spent nuclear fuel

Commercial waste and spent fuel packaging program. Annual report

International Nuclear Information System (INIS)

Hakl, A.R.

1981-10-01

This document is a report of activities performed by Westinghouse Advanced Energy Systems Division - Nevada Operations in meeting subtask objectives described in the Nevada Nuclear Waste Storage Investigations (NNWSI) Project Plan and revised planning documentation for Fiscal Year (FY) 1981. Major activities included: completion of the first fuel exchange in the Spent Fuel Test - Climax program; plasma arc welder development; modification and qualification of a canister cutter; installation, and activation of a remote area monitor, constant air monitor and an alpha/beta/gamma counting system; qualification of grapples required to handle pressurized water reactor or boiling water reactor fuel and high level waste (HLW) logs; data acquisition from the 3 kilowatt soil temperature test, 2 kw fuel temperature test, and 2 kw drywell test; calorimetry of the fuel assembly used in the fuel temperature test; evaluation of moisture accumulation in the drywells and recommendations for proposed changes; revision of safety assessment document to include HLW log operations; preparation of quality assurance plan and procedures; development and qualification of all equipment and procedures to receive, handle and encapsulate both the HLW log and spent fuel for the basalt waste isolation program/near surface test facility program; preliminary studies of both the requirements to perform waste packaging for the test and evaluation facility and a cask storage program for the DOE Interim Spent Fuel Management program; and remote handling operations on radioactive source calibration in support of other contractors

Review and problem definition of water/rock reactions associated with injection of spent geothermal fluids from a geothermal plant into aquifers

Energy Technology Data Exchange (ETDEWEB)

Elders, W.A.

1986-07-01

Among the technical problems faced by the burgeoning geothermal industry is the disposal of spent fluids from power plants. Except in unusual circumstances the normal practice, especially in the USA, is to pump these spent fluids into injection wells to prevent contamination of surface waters, and possibly in some cases, to reduce pressure drawdown in the producing aquifers. This report is a survey of experience in geothermal injection, emphasizing geochemical problems, and a discussion of approaches to their possible mitigation. The extraction of enthalpy from geothermal fluid in power plants may cause solutions to be strongly supersaturated in various dissolved components such as silica, carbonates, sulfates, and sulfides. Injection of such supersaturated solutions into disposal wells has the potential to cause scaling in the well bores and plugging of the aquifers, leading to loss of injectivity. Various aspects of the geochemistry of geothermal brines and their potential for mineral formation are discussed, drawing upon a literature survey. Experience of brine treatment and handling, and the economics of mineral extraction are also addressed in this report. Finally suggestions are made on future needs for possible experimental, field and theoretical studies to avoid or control mineral scaling.

Modelling of the spent fuel heat-up in the spent fuel pools using one-dimensional system codes and CFD codes

Energy Technology Data Exchange (ETDEWEB)

Grazevicius, Audrius; Kaliatka, Algirdas [Lithuanian Energy Institute, Kaunas (Lithuania). Lab. of Nuclear Installation Safety

2017-07-15

The main functions of spent fuel pools are to remove the residual heat from spent fuel assemblies and to perform the function of biological shielding. In the case of loss of heat removal from spent fuel pool, the fuel rods and pool water temperatures would increase continuously. After the saturated temperature is reached, due to evaporation of water the pool water level would drop, eventually causing the uncover of spent fuel assemblies, fuel overheating and fuel rods failure. This paper presents an analysis of loss of heat removal accident in spent fuel pool of BWR 4 and a comparison of two different modelling approaches. The one-dimensional system thermal-hydraulic computer code RELAP5 and CFD tool ANSYS Fluent were used for the analysis. The results are similar, but the local effects cannot be simulated using a one-dimensional code. The ANSYS Fluent calculation demonstrated that this three-dimensional treatment allows to avoid the need for many one-dimensional modelling assumptions in the pool modelling and enables to reduce the uncertainties associated with natural circulation flow calculation.

Water-Based Pressure Sensitive Paint

Science.gov (United States)

Oglesby, Donald M.; Ingram, JoAnne L.; Jordan, Jeffrey D.; Watkins, A. Neal; Leighty, Bradley D.

2004-01-01

Preparation and performance of a water-based pressure sensitive paint (PSP) is described. A water emulsion of an oxygen permeable polymer and a platinum porphyrin type luminescent compound were dispersed in a water matrix to produce a PSP that performs well without the use of volatile, toxic solvents. The primary advantages of this PSP are reduced contamination of wind tunnels in which it is used, lower health risk to its users, and easier cleanup and disposal. This also represents a cost reduction by eliminating the need for elaborate ventilation and user protection during application. The water-based PSP described has all the characteristics associated with water-based paints (low toxicity, very low volatile organic chemicals, and easy water cleanup) but also has high performance as a global pressure sensor for PSP measurements in wind tunnels. The use of a water-based PSP virtually eliminates the toxic fumes associated with the application of PSPs to a model in wind tunnels.

High-pressure water facility

Science.gov (United States)

2006-01-01

NASA Test Operations Group employees, from left, Todd Pearson, Tim Delcuze and Rodney Wilkinson maintain a water pump in Stennis Space Center's high-pressure water facility. The three were part of a group of employees who rode out Hurricane Katrina at the facility and helped protect NASA's rocket engine test complex.

SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions

Energy Technology Data Exchange (ETDEWEB)

Radulescu, Georgeta [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL

2010-03-01

The purpose of this calculation report is to document the comparison to measurement of the isotopic concentrations for pressurized water reactor (PWR) spent nuclear fuel determined with the Standardized Computer Analysis for Licensing Evaluation (SCALE) 5.1 (Ref. ) epletion calculation method. Specifically, the depletion computer code and the cross-section library being evaluated are the twodimensional (2-D) transport and depletion module, TRITON/NEWT,2, 3 and the 44GROUPNDF5 (Ref. 4) cross-section library, respectively, in the SCALE .1 code system.

Impacts of the use of spent nuclear fuel burnup credit on DOE advanced technology legal weight truck cask GA-4 fleet size

International Nuclear Information System (INIS)

Mobasheran, A.S.; Boshoven, J.; Lake, B.

1995-01-01

The object of this paper is to study the impact of full and partial spent fuel burnup credit on the capacity of the Legal Weight Truck Spent Fuel Shipping Cask (GA-4) and to determine the numbers of additional spent fuel assemblies which could be accommodated as a result. The scope of the study comprised performing nuclear criticality safety scoping calculations using the SCALE-PC software package and the 1993 spent fuel database to determine logistics for number of spent fuel assemblies to be shipped. The results of the study indicate that more capacity than 2 or 3 pressurized water reactor assemblies could be gained for GA-4 casks when burnup credit is considered. Reduction in GA-4 fleet size and number of shipments are expected to result from the acceptance of spent fuel burnup credit

Long term wet spent nuclear fuel storage

International Nuclear Information System (INIS)

1987-04-01

The meeting showed that there is continuing confidence in the use of wet storage for spent nuclear fuel and that long-term wet storage of fuel clad in zirconium alloys can be readily achieved. The importance of maintaining good water chemistry has been identified. The long-term wet storage behaviour of sensitized stainless steel clad fuel involves, as yet, some uncertainties. However, great reliance will be placed on long-term wet storage of spent fuel into the future. The following topics were treated to some extent: Oxidation of the external surface of fuel clad, rod consolidation, radiation protection, optimum methods of treating spent fuel storage water, physical radiation effects, and the behaviour of spent fuel assemblies of long-term wet storage conditions. A number of papers on national experience are included

PWR and BWR spent fuel assembly gamma spectra measurements

Energy Technology Data Exchange (ETDEWEB)

Vaccaro, S. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Tobin, S.J.; Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Hu, J. [Oak Ridge National Laboratory, Oak Ridge (United States); Schwalbach, P. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company (SKB) (Sweden); Trellue, H.; Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States)

2016-10-11

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of {sup 137}Cs, {sup 154}Eu, and {sup 134}Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

Investigation of the thermal performance of a vertical two-phase closed thermosyphon as a passive cooling system for a nuclear reactor spent fuel storage pool

Energy Technology Data Exchange (ETDEWEB)

Kusuma, Mukhsinun Hadi; Putra, Nandy; Imawan, Ficky Augusta [Heat Transfer Laboratory, Department of Mechanical Engineering Universitas Indonesia, Kampus (Indonesia); Antariksawan, Anhar Riza [Centre for Nuclear Reactor Safety and Technology, National Nuclear Energy Agency of Indonesia (BATAN), Kawasan Puspiptek Serpong (Indonesia)

2017-04-15

The decay heat that is produced by nuclear reactor spent fuel must be cooled in a spent fuel storage pool. A wickless heat pipe or a vertical two-phase closed thermosyphon (TPCT) is used to remove this decay heat. The objective of this research is to investigate the thermal performance of a prototype model for a large-scale vertical TPCT as a passive cooling system for a nuclear research reactor spent fuel storage pool. An experimental investigation and numerical simulation using RELAP5/MOD 3.2 were used to investigate the TPCT thermal performance. The effects of the initial pressure, filling ratio, and heat load were analyzed. Demineralized water was used as the TPCT working fluid. The cooled water was circulated in the water jacket as a cooling system. The experimental results show that the best thermal performance was obtained at a thermal resistance of 0.22°C/W, the lowest initial pressure, a filling ratio of 60%, and a high evaporator heat load. The simulation model that was experimentally validated showed a pattern and trend line similar to those of the experiment and can be used to predict the heat transfer phenomena of TPCT with varying inputs.

Pressure pressure-balanced pH sensing system for high temperature and high pressure water

International Nuclear Information System (INIS)

Tachibana, Koji

1995-01-01

As for the pH measurement system for high temperature, high pressure water, there have been the circumstances that first the reference electrodes for monitoring corrosion potential were developed, and subsequently, it was developed for the purpose of maintaining the soundness of metallic materials in high temperature, high pressure water in nuclear power generation. In the process of developing the reference electrodes for high temperature water, it was clarified that the occurrence of stress corrosion cracking in BWRs is closely related to the corrosion potential determined by dissolved oxygen concentration. As the types of pH electrodes, there are metal-hydrogen electrodes, glass electrodes, ZrO 2 diaphragm electrodes and TiO 2 semiconductor electrodes. The principle of pH measurement using ZrO 2 diaphragms is explained. The pH measuring system is composed of YSZ element, pressure-balanced type external reference electrode, pressure balancer and compressed air vessel. The stability and pH response of YSZ elements are reported. (K.I.)

Managing water pressure for water savings in developing countries

African Journals Online (AJOL)

2014-03-03

Mar 3, 2014 ... effort into providing customers with a reliable level of service, often via poor water ... budgets. There are many factors contributing to water losses in water .... given relationship does not reflect the impact of pressure on.

Pressurized water-reactor feedwater piping response to water hammer

International Nuclear Information System (INIS)

Arthur, D.

1978-03-01

The nuclear power industry is interested in steam-generator water hammer because it has damaged the piping and components at pressurized water reactors (PWRs). Water hammer arises when rapid steam condensation in the steam-generator feedwater inlet of a PWR causes depressurization, water-slug acceleration, and slug impact at the nearest pipe elbow. The resulting pressure pulse causes the pipe system to shake, sometimes violently. The objective of this study is to evaluate the potential structural effects of steam-generator water hammer on feedwater piping. This was accomplished by finite-element computation of the response of two sections of a typical feedwater pipe system to four representative water-hammer pulses. All four pulses produced high shear and bending stresses in both sections of pipe. Maximum calculated pipe stresses varied because the sections had different characteristics and were sensitive to boundary-condition modeling

Storage of LWR spent fuel in air: Volume 1: Design and operation of a spent fuel oxidation test facility

International Nuclear Information System (INIS)

Thornhill, C.K.; Campbell, T.K.; Thornhill, R.E.

1988-12-01

This report describes the design and operation and technical accomplishments of a spent-fuel oxidation test facility at the Pacific Northwest Laboratory. The objective of the experiments conducted in this facility was to develop a data base for determining spent-fuel dry storage temperature limits by characterizing the oxidation behavior of light-water reactor (LWR) spent fuels in air. These data are needed to support licensing of dry storage in air as an alternative to spent-fuel storage in water pools. They are to be used to develop and validate predictive models of spent-fuel behavior during dry air storage in an Independent Spent Fuel Storage Installation (ISFSI). The present licensed alternative to pool storage of spent fuel is dry storage in an inert gas environment, which is called inerted dry storage (IDS). Licensed air storage, however, would not require monitoring for maintenance of an inert-gas environment (which IDS requires) but does require the development of allowable temperature limits below which UO 2 oxidation in breached fuel rods would not become a problem. Scoping tests at PNL with nonirradiated UO 2 pellets and spent-fuel fragment specimens identified the need for a statistically designed test matrix with test temperatures bounding anticipated maximum acceptable air-storage temperatures. This facility was designed and operated to satisfy that need. 7 refs

FY13 Summary Report on the Augmentation of the Spent Fuel Composition Dataset for Nuclear Forensics: SFCOMPO/NF

Energy Technology Data Exchange (ETDEWEB)

Brady Raap, Michaele C.; Lyons, Jennifer A.; Collins, Brian A.; Livingston, James V.

2014-03-31

This report documents the FY13 efforts to enhance a dataset of spent nuclear fuel isotopic composition data for use in developing intrinsic signatures for nuclear forensics. A review and collection of data from the open literature was performed in FY10. In FY11, the Spent Fuel COMPOsition (SFCOMPO) excel-based dataset for nuclear forensics (NF), SFCOMPO/NF was established and measured data for graphite production reactors, Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs) were added to the dataset and expanded to include a consistent set of data simulated by calculations. A test was performed to determine whether the SFCOMPO/NF dataset will be useful for the analysis and identification of reactor types from isotopic ratios observed in interdicted samples.

Possible effects of UO2 oxidation on light water reactor spent fuel performance in long-term geologic disposal

International Nuclear Information System (INIS)

Almassy, M.Y.; Woodley, R.E.

1982-08-01

Disposal of spent nuclear fuel in a conventionally mined geologic formation is the nearest-term option for permanently isolating radionuclides from the biosphere. Because irradiated uranium dioxide (UO 2 ) fuel pellets retain 95 to 99% of the radionuclides generated during normal light water reactor operation, they may represent a significant barrier to radionuclide release. This document presents a technical assessment of published literature representing the current level of understanding of spent fuel characteristics and conditions that may degrade pellet integrity during a geologic disposal sequence. A significant deterioration mechanism is spent UO 2 oxidation with possible consequences identified as fission gas release, rod diameter increases, cladding breach extension, and release of solid fuel particles containing radionuclides. Areas requiring further study to support development of a comprehensive spent fuel performance prediction model are highlighted. A program and preliminary schedule to obtain the information needed to develop model correlations are also presented

Behavior of spent nuclear fuel in water pool storage

International Nuclear Information System (INIS)

Johnson, A.B. Jr.

1977-09-01

Storage of irradiated nuclear fuel in water pools (basins) has been standard practice since nuclear reactors first began operation approximately 34 years ago. Pool storage is the starting point for all other fuel storage candidate processes and is a candidate for extended interim fuel storage until policy questions regarding reprocessing and ultimate disposal have been resolved. This report assesses the current performance of nuclear fuel in pool storage, the range of storage conditions, and the prospects for extending residence times. The assessment is based on visits to five U.S. and Canadian fuel storage sites, representing nine storage pools, and on discussions with operators of an additional 21 storage pools. Spent fuel storage experience from British pools at Winfrith and Windscale and from a German pool at Karlsruhe (WAK) also is summarized

Quality of water from the pool, original containers and aluminum drums used for storage of spent nuclear fuel

International Nuclear Information System (INIS)

Idjakovic, Z.; Milonjic, S.; Cupic, S.

2001-01-01

Results of chemical analyses of water from the pool, including original containers and aluminium drums, for storage of spent nuclear fuel of the research reactor RA at the VINCA Institute and a short survey of the water properties from similar pools of other countries are presented in the paper. (author)

CIRFT Data Update and Data Analyses for Spent Nuclear Fuel Vibration Reliability Study

Energy Technology Data Exchange (ETDEWEB)

Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2018-01-01

The objective of this research is to collect experimental data on spent nuclear fuel (SNF) from pressurized water reactors (PWRs), including the H. B. Robinson Nuclear Power Station (HBR), Catawba Nuclear Station, North Anna Nuclear Power Station (NA), and the Limerick Nuclear Power Station (LMK) boiling water reactor (BWR). Data will be collected under simulated transportation environments using the cyclic integrated reversible-bending fatigue tester (CIRFT), an enabling hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL). These data will be used to support ongoing SNF modeling activities and to address regulatory issues associated with SNF transport.

Final Generic Environmental Impact Statement. Handling and storage of spent light water power reactor fuel. Volume 1. Executive summary and text

International Nuclear Information System (INIS)

1979-08-01

The Generic Environmental Impact Statement on spent fuel storage was prepared by the Nuclear Regulatory Commission staff in response to a directive from the Commissioners published in the Federal Register, September 16, 1975 (40 FR 42801). The Commission directed the staff to analyze alternatives for the handling and storage of spent light water power reactor fuel with particular emphasis on developing long range policy. Accordingly, the scope of this statement examines alternative methods of spent fuel storage as well as the possible restriction or termination of the generation of spent fuel through nuclear power plant shutdown. Volume 1 includes the executive summary and the text

Radiological characteristics of light-water reactor spent fuel: A literature survey of experimental data

International Nuclear Information System (INIS)

Roddy, J.W.; Mailen, J.C.

1987-12-01

This survey brings together the experimentally determined light-water reactor spent fuel data comprising radionuclide composition, decay heat, and photon and neutron generation rates as identified in a literature survey. Many citations compare these data with values calculated using a radionuclide generation and depletion computer code, ORIGEN, and these comparisons have been included. ORIGEN is a widely recognized method for estimating the actinide, fission product, and activation product contents of irradiated reactor fuel, as well as the resulting heat generation and radiation levels. These estimates are used as source terms in safety evaluations of operating reactors, for evaluation of fuel behavior and regulation of the at-reactor storage, for transportation studies, and for evaluation of the ultimate geologic storage of spent fuel. 82 refs., 4 figs., 17 tabs

Development of a water boil-off spent-fuel calorimeter system

International Nuclear Information System (INIS)

Creer, J.M.; Shupe, J.W. Jr.

1981-05-01

A calorimeter system was developed to measure decay heat generation rates of unmodified spent fuel assemblies from commercial nuclear reactors. The system was designed, fabricated, and successfully tested using the following specifications: capacity of one BWR or PWR spent fuel assembly; decay heat generation range 0.1 to 2.5 kW; measurement time of < 12 h; and an accuracy of +-10% or better. The system was acceptance tested using a dc reference heater to simulate spent fuel assembly heat generation rates. Results of these tests indicated that the system could be used to measure heat generation rates between 0.5 and 2.5 kW within +- 5%. Measurements of heat generation rates of approx. 0.1 kW were obtained within +- 15%. The calorimeter system has the potential to permit measurements of heat generation rates of spent fuel assemblies and other devices in the 12- to 14-kW range. Results of calorimetry of a Turkey Point spent fuel assembly indicated that the assembly was generating approx. 1.55 kW

Spent fuel surveillance and monitoring methods

International Nuclear Information System (INIS)

1988-05-01

The Technical Committee Meeting on ''Spent Fuel Surveillance and Monitoring Methods'' (27-30 October 1987) has been organized in accordance with recommendations of the International Standing Advisory Group on Spent Fuel Management during its second meeting in 1986. The aim of the meeting was to discuss the above questions with emphasis on current design and operation criteria, safety principles and licensing requirements and procedures in order to prevent: inadvertent criticality, undue radiation exposure, unacceptable release of radioactivity as well as control for loss of storage pool water, crud impact, water chemistry, distribution and behaviour of particulates in cooling water, oxidation of intact and failed fuel rods as a function of temperature and burnup; distribution of radiation and temperature through dry cask wall, monitoring of leakages from pools and gas escapes from dry storage facilities, periodical integrity tests of the containment barriers, responsibilities of organizations for the required operation, structure, staff and subordination, etc. The presentations of the Meeting were divided into two sessions: Spent fuel surveillance programmes and practice in Member States (4 papers); Experimental methods developed in support of spent fuel surveillance programmes (5 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

The burnup dependence of light water reactor spent fuel oxidation

International Nuclear Information System (INIS)

Hanson, B.D.

1998-07-01

Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO 2 is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO 2 to higher oxides. The oxidation of UO 2 has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO 2 oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO 2 to UO 2.4 was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO 2.4 to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO 2 oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO 2 and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies associated with spent fuel oxidation (Section 5)

Water cooled static pressure probe

Science.gov (United States)

Lagen, Nicholas T. (Inventor); Eves, John W. (Inventor); Reece, Garland D. (Inventor); Geissinger, Steve L. (Inventor)

1991-01-01

An improved static pressure probe containing a water cooling mechanism is disclosed. This probe has a hollow interior containing a central coolant tube and multiple individual pressure measurement tubes connected to holes placed on the exterior. Coolant from the central tube symmetrically immerses the interior of the probe, allowing it to sustain high temperature (in the region of 2500 F) supersonic jet flow indefinitely, while still recording accurate pressure data. The coolant exits the probe body by way of a reservoir attached to the aft of the probe. The pressure measurement tubes are joined to a single, larger manifold in the reservoir. This manifold is attached to a pressure transducer that records the average static pressure.

Effects of environments on spent fuel

International Nuclear Information System (INIS)

Funk, C.W.; Jacobson, L.D.; Menon, M.N.

1979-07-01

This report describes the influence of water storage environment and transportation on spent light water reactor (LWR) fuel assemblies. It also estimates the storage duration and capacity requirements for several assumed scenarios

Probable leaching mechanisms for spent fuel

International Nuclear Information System (INIS)

Wang, R.; Katayama, Y.B.

1981-01-01

At the Pacific Northwest Laboratory, researchers in the Waste/Rock Interaction Technology Program are studying spent fuel as a possible waste form for the Office of Nuclear Waste Isolation. This paper presents probable leaching mechanisms for spent fuel and discusses current progress in identifying and understanding the leaching process. During the past year, experiments were begun to study the complex leaching mechanism of spent fuel. The initial work in this investigation was done with UO 2 , which provided the most information possible on the behavior of the spent-fuel matrix without encountering the very high radiation levels associated with spent fuel. Both single-crystal and polycrystalline UO 2 samples were used for this study, and techniques applicable to remote experimentation in a hot cell are being developed. The effects of radiation are being studied in terms of radiolysis of water and surface activation of the UO 2 . Dissolution behavior and kinetics of UO 2 were also investigated by electrochemical measurement techniques. These data will be correlated with those acquired when spent fuel is tested in a hot cell. Oxidation effects represent a major area of concern in evaluating the stability of spent fuel. Dissolution of UO 2 is greatly increased in an oxidizing solution because the dissolution is then controlled by the formation of hexavalent uranium. In solutions containing very low oxygen levels (i.e., reducing solutions), oxidation-induced dissolution may be possible via a previously oxidized surface, through exposure to air during storage, or by local oxidants such as O 2 and H 2 O 2 produced from radiolysis of water and radiation-activated UO 2 surfaces. The effects of oxidation not only increase the dissolution rate, but could lead to the disintegration of spent fuel into fine fragments

Retrofitting Trojan Nuclear Plant's spent resin transfer system

International Nuclear Information System (INIS)

Pierce, R.E.

1979-01-01

The spent resin slurry transport system at the Trojan Nuclear Plant operated by Portland General Electric Company is one of the most advanced systems of its type in the nuclear industry today. The new system affords the plant's operators safe remote sonic indication for spent resin and cover water levels, manual remote dewatering and watering capability to establish desirable resin-to-water volumetric ratios, reliable non-mechanical resin agitation utilizing fixed spargers, and controllable process flow utilizing a variable speed recessed impeller pump

Water Pressure Distribution on a Flying Boat Hull

Science.gov (United States)

Thompson, F L

1931-01-01

This is the third in a series of investigations of the water pressures on seaplane floats and hulls, and completes the present program. It consisted of determining the water pressures and accelerations on a Curtiss H-16 flying boat during landing and taxiing maneuvers in smooth and rough water.

Nanobiotechnology for the production of biofuels from spent tea ...

African Journals Online (AJOL)

Bioenergy is the only alternative and cheap source of energy which can be made easily available to the world. The present experiment included three steps for the conversion of spent tea (Camellia sinensis) into biofuels. In the first step, spent tea was gasified using Co nano catalyst at 300°C and atmospheric pressure.

Where Did the Water Go?: Boyle's Law and Pressurized Diaphragm Water Tanks

Science.gov (United States)

Brimhall, James; Naga, Sundar

2007-01-01

Many homes use pressurized diaphragm tanks for storage of water pumped from an underground well. These tanks are very carefully constructed to have separate internal chambers for the storage of water and for the air that provides the pressure. One might expect that the amount of water available for use from, for example, a 50-gallon tank would be…

Pressure swing adsorption cycle for natural gas pretreatment for liquefaction

Energy Technology Data Exchange (ETDEWEB)

Mann, R.E.

1984-01-10

An improved apparatus and method for a regeneration of a solid adsorbent used to pretreat a gas before liquefaction. The spent adsorbent is subjected to a two-step regenerative process, in the first of which the adsorbent is subjected to a low pressure produced by the use of mechanical vacuum pumps. When the pressure of the atmosphere in contact with the adsorbent has been reduced to a level sufficiently low to insure that the gas will flow under laminar rather than viscous conditions, the adsorbent is exposed to the action of a cryoplate condenser maintained at a sufficiently low temperature to cause any molecules of water which impinge thereon to condense and freeze, thereby reducing the partial pressure of water vapor within the chamber. The reduced partial pressure of the water vapor in turn causes adsorbed water on the adsorbent to be desorbed, thereby effectively removing the water from the adsorbent and depositing it as solid ice on the cryoplate condenser.

The effect of fuel burnup and dispersed water intrusion on the criticality of spent high-level nuclear fuel in a geologic repository

International Nuclear Information System (INIS)

Culbreth, W.G.; Zielinski, P.R.

1994-01-01

Studies of the spent fuel waste package have been conducted through the use of a Monte-Carlo neutron simulation program to determine the ability of the fuel to sustain a chain reaction. These studies have included fuel burnup and the effect of water mists on criticality. Results were compared with previous studies. In many criticality studies of spent fuel waste packages, fresh fuel with an enrichment as high as 4.5% is used as the conservative (worst) case. The actual spent fuel has a certain amount of burnup that decreases the concentration of fissile uranium and increases the amount of radionuclides present. The LWR Radiological Data Base from OCRWM has been used to determine the relative radionuclide ratios and KENO 5.1 was used to calculate values of the effective multiplication factor, k eff . Spent fuel is not capable of sustaining a chain reaction unless a suitable moderator, such as water, is present. A completely flooded container has been treated as the worst case for criticality. Results of a previous report that demonstrated that k eff actually peaked at a water-to-mixture ratio of 13% were analyzed for validity. In the present study, these results did not occur in the SCP waste package container

Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods

International Nuclear Information System (INIS)

Johnson, G.L.

1988-09-01

Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. One such package would store lightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97/degree/C and whether the cladding of the stored spent fuel ever exceeds 350/degree/C. Limiting the borehole to temperatures of 97/degree/C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350/degree/C cladding limit minimizes the possibility of creep-related failure in the spent fuel rod cladding. For a series of packages stored in a 8 x 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97/degree/C for the full 1000-yr analysis period

Final environmental impact statement: US Spent Fuel Policy. Charge for spent fuel storage

International Nuclear Information System (INIS)

1980-05-01

The United States Government policy relating to nuclear fuel reprocessing, which was announced by President Carter on April 7, 1977, provides for an indefinite deferral of reprocessing, and thus commits light water reactor (LWR) plants to a once-through fuel cycle during that indefinite period. In a subsequent action implementing that policy, the Department of Energy (DOE) on October 18, 1977 announced a spent fuel policy which would enable domestic, and on a selective basis, foreign utilities to deliver spent fuel to the US Government for interim storage and final geologic disposal, and pay the Government a fee for such services. This volume addresses itself to whether the fee charged for these services, by its level or its structure, would have any effect on the environmental impacts of implementing the Spent Fuel Policy itself. This volume thus analyzes the fee and various alternatives to determine the interaction between the fee and the degree of participation by domestic utilities and foreign countries in the proposed spent fuel program for implementing the Spent Fuel Policy. It also analyzes the effect, if any, of the fee on the growth of nuclear power

Spent fuel's behavior under dynamic drip tests

International Nuclear Information System (INIS)

Finn, P.A.; Buck, E.C.; Hoh, J.C.; Bates, J.K.

1995-01-01

In the potential repository at Yucca Mountain, failure of the waste package container and the cladding of the spent nuclear fuel would expose the fuel to water under oxidizing conditions. To simulate the release behavior of radionuclides from spent fuel, dynamic drip and vapor tests with spent nuclear fuel have been ongoing for 2.5 years. Rapid alteration of the spent fuel has been noted with concurrent release of radionuclides. Colloidal species containing americium and plutonium have been found in the leachate. This observation suggests that colloidal transport of radionuclides should be included in the performance assessment of a potential repository

A Preliminary Study on the Reuse of the Recovered Uranium from the Spent CANDU Fuel Using Pyroprocessing

International Nuclear Information System (INIS)

Park, C. J.; Na, S. H.; Yang, J. H.; Kang, K. H.; Lee, J. W.

2009-01-01

During the pyroprocessing, most of the uranium is gathered in metallic form around a solid cathode during an electro-refining process, which is composed of about 94 weight percent of the spent fuel. In the previous study, a feasibility study has been done to reuse the recovered uranium for the CANDU reactor fuel following the traditional DUPIC (direct use of spent pressurized water reactor fuel into CANDU reactor) fuel fabrication process. However, the weight percent of U-235 in the recovered uranium is about 1 wt% and it is sufficiently re-utilized in a heavy water reactor which uses a natural uranium fuel. The reuse of recovered uranium will bring not only a huge economic profit and saving of uranium resources but also an alleviation of the burden on the management and the disposal of the spent fuel. The research on recycling of recovered uranium was carried out 10 years ago and most of the recovered uranium was assumed to be imported from abroad at that time. The preliminary results showed there is the sufficient possibility to recycle recovered uranium in terms of a reactor's characteristics as well as the fuel performance. However, the spent CANDU fuel is another issue in the storage and disposal problem. At present, most countries are considering that the spent CANDU fuel is disposed directly due to the low enrichment (∼0.5 wt%) of the discharge fissile content and lots of fission products. If mixing the spent CANDU fuel and the spent PWR fuel, the estimated uranium fissile enrichment will be about 0.6 wt% ∼ 1.0 wt% depending on the mixing ratio, which is sufficiently reusable in a CANDU reactor. Therefore, this paper deals with a feasibility study on the recovered uranium of the mixed spent fuel from the pyroprocessing. With the various mixing ratios between the PWR spent fuel and the CANDU spent fuel, a reactor characteristics including the safety parameters of the CANDU reactor was evaluated

Preliminary Calculation on a Spent Fuel Pool Accident using GOTHIC

Energy Technology Data Exchange (ETDEWEB)

Park, Jaehwan; Choi, Yu Jung; Hong, Tae Hyub; Kim, Hyeong-Taek [KHNP-CRI, Daejeon (Korea, Republic of)

2015-10-15

The probability of an accident happening at the spent fuel pool was believed to be quite low until the 2011 Fukushima accident occurred. Notably, large amount of spent fuel are normally stored in the spent fuel pool for a long time compared to the amount of fuel in the reactor core and the total heat released from the spent fuel is high enough to boil the water of the spent fuel pool when the cooling system does not operate. In addition, the enrichment and the burnup of the fuel have both increased in the past decade and heat generation from the spent fuel thereby has also increased. The failure of the cooling system at the spent fuel pool (hereafter, a loss-of-cooling accident) is one of the principal hypothetical causes of an accident that could occur at the spent fuel pool. In this paper, the preliminary calculation of a loss-of-cooling accident was performed. In this paper, the preliminary calculation of a loss-of cooling accident was performed with GOTHIC. The calculation results show boiling away of water in the spent fuel pool due to the loss-of-cooling accident and similar thermal performance of the spent fuel pool with previous research results.

Measurement of plutonium in spent nuclear fuel by self-induced x-ray fluorescence

Energy Technology Data Exchange (ETDEWEB)

Hoover, Andrew S [Los Alamos National Laboratory; Rudy, Cliff R [Los Alamos National Laboratory; Tobin, Steve J [Los Alamos National Laboratory; Charlton, William S [Los Alamos National Laboratory; Stafford, A [TEXAS A& M; Strohmeyer, D [TEXAS A& M; Saavadra, S [ORNL

2009-01-01

Direct measurement of the plutonium content in spent nuclear fuel is a challenging problem in non-destructive assay. The very high gamma-ray flux from fission product isotopes overwhelms the weaker gamma-ray emissions from plutonium and uranium, making passive gamma-ray measurements impossible. However, the intense fission product radiation is effective at exciting plutonium and uranium atoms, resulting in subsequent fluorescence X-ray emission. K-shell X-rays in the 100 keV energy range can escape the fuel and cladding, providing a direct signal from uranium and plutonium that can be measured with a standard germanium detector. The measured plutonium to uranium elemental ratio can be used to compute the plutonium content of the fuel. The technique can potentially provide a passive, non-destructive assay tool for determining plutonium content in spent fuel. In this paper, we discuss recent non-destructive measurements of plutonium X-ray fluorescence (XRF) signatures from pressurized water reactor spent fuel rods. We also discuss how emerging new technologies, like very high energy resolution microcalorimeter detectors, might be applied to XRF measurements.

AP1000{sup R} nuclear power plant safety overview for spent fuel cooling

Energy Technology Data Exchange (ETDEWEB)

Gorgemans, J.; Mulhollem, L.; Glavin, J.; Pfister, A.; Conway, L.; Schulz, T.; Oriani, L.; Cummins, E.; Winters, J. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

The AP1000{sup R} plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and costs. The AP1000 design uses passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems such as AC power, component cooling water, service water or HVAC. Furthermore, these passive features 'fail safe' during a non-LOCA event such that DC power and instrumentation are not required. The AP1000 also has simple, active, defense-in-depth systems to support normal plant operations. These active systems provide the first level of defense against more probable events and they provide investment protection, reduce the demands on the passive features and support the probabilistic risk assessment. The AP1000 passive safety approach allows the plant to achieve and maintain safe shutdown in case of an accident for 72 hours without operator action, meeting the expectations provided in the U.S. Utility Requirement Document and the European Utility Requirements for passive plants. Limited operator actions are required to maintain safe conditions in the spent fuel pool via passive means. In line with the AP1000 approach to safety described above, the AP1000 plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for design-basis events and beyond design-basis accidents. During normal and abnormal conditions, defense-in-depth and other systems provide highly reliable spent fuel pool cooling. They rely on off-site AC power or the on-site standby diesel generators. For unlikely design basis events with an extended loss of AC power (i.e., station blackout) or loss of heat sink or both, spent fuel cooling can still be provided indefinitely: - Passive systems, requiring minimal or no operator actions, are sufficient for at least 72 hours under all

Pressurized water reactor flow arrangement

International Nuclear Information System (INIS)

Gibbons, J.F.; Knapp, R.W.

1980-01-01

A flow path is provided for cooling the control rods of a pressurized water reactor. According to this scheme, a small amount of cooling water enters the control rod guide tubes from the top and passes downwards through the tubes before rejoining the main coolant flow and passing through the reactor core. (LL)

Solubility and physical properties of sugars in pressurized water

International Nuclear Information System (INIS)

Saldaña, Marleny D.A.; Alvarez, Víctor H.; Haldar, Anupam

2012-01-01

Highlights: ► Sugar solubility in pressurized water and density at high pressures were measured. ► Glucose solubility was higher than that of lactose as predicted by their σ-profiles. ► Sugar aqueous solubility decreased with an increase in pressure from 15 to 120 bar. ► Aqueous glucose molecular packing shows high sensitivity to pressure. ► The COSMO-SAC model qualitatively predicted the sugar solubility data. - Abstract: In this study, the solubility, density, and refractive index of glucose and lactose in water as a function of temperature were measured. For solubility of sugars in pressurized water, experimental data were obtained at pressures of (15 to 120) bar and temperatures of (373 to 433) K using a dynamic flow high pressure system. Density data for aqueous sugar solutions were obtained at pressures of (1 to 300) bar and temperatures of (298 to 343) K. The refractive index of aqueous sugar solutions was obtained at 293 K and atmospheric pressure. Activity coefficient models, Van Laar and the Conductor-like Screening Model-Segment Activity Coefficient (COSMO-SAC), were used to fit and predict the experimental solubility data, respectively. The results obtained showed that the solubility of both sugars in pressurized water increase with an increase in temperature. However, with the increase of pressure from 15 bar to 120 bar, the solubility of both sugars in pressurized water decreased. The Van Laar model fit the experimental aqueous solubility data with deviations lower than 13 and 53% for glucose and lactose, respectively. The COSMO-SAC model predicted qualitatively the aqueous solubility of these sugars.

Some factors to consider in handling and storing spent fuel

International Nuclear Information System (INIS)

Bailey, W.J.

1985-11-01

This report includes information from various studies performed under the Wet Storage Task of the Behavior of Spent Fuel in Storage Project of the Commercial Spent Fuel Management (CSFM) Program at Pacific Northwest Laboratory. Wet storage experience has been summarized earlier in several other reports. This report summarizes pertinent items noted during FY 1985 concerning recent developments in the handling and storage of spent fuel and associated considerations. The subjects discussed include recent publications, findings, and developments associated with: (1) storage of water reactor spent fuel in water pools, (2) extended-burnup fuel, (3) fuel assembly reconstitution and reinsertion, (4) rod consolidation, (5) variations in the US Nuclear Regulatory Commission's definition of failed fuel, (6) detection of failed fuel rods, and (7) extended integrity of spent fuel. A list of pertinent publications is included

Experience and prospects of WWER-1000 reactor spent fuel transport

International Nuclear Information System (INIS)

Kondratyev, A.N.; Yershov, V.N.; Kozlov, Yu.V.; Kosarev, Yu.A.; Ilyin, Yu.V.; Pavlov, M.S.

1989-01-01

The paper deals with the USSR experience in shipping the commercial WWER-1000 reactor spent fuel in TK-10 and TK-13 casks. The cask designs, their basic characteristics and the WWER-1000 spent fuel features are described. An example of calculational/experimental approach in the design of a basket (one of the most important components) for spent fuel assembly (SFA) accommodation in a cask is given. The main problems of future development works are presented in brief. A concept of development of nuclear power industry with the closed fuel cycle is assumed in the Soviet Union, hence the spent nuclear fuel is to be transported from NPPs to reprocessing plants. To transport WWER-1000 spent fuel, the casks of two types were developed. These are: a pilot TK-10 cask of 3t capacity in fuel; a commercial TK-13 cask of ∼6t capacity in fuel. The pilot TK-10 cask is thick-walled (360mm) cylindrical vessel manufactured of steel shells and a bottom welded to each other. The material of the body is carbon steel. There is a steel jacket on the outer side of the cask body and at 120 mm distance off the bottom. On its cylindrical part between the jacket and the body there are T-shaped circular ribs acting as shock-absorbers. The space between the jacket and the body is filled with ethylene glycol solution of 65 degree C crystallization temperature, which functions as a neutron shielding. The TK-10 cask coolant is water or air (nitrogen) at minor excess pressure resulted from FA heatup after the cask sealing

High-pressure-induced water penetration into 3-isopropylmalate dehydrogenase

International Nuclear Information System (INIS)

Nagae, Takayuki; Kawamura, Takashi; Chavas, Leonard M. G.; Niwa, Ken; Hasegawa, Masashi; Kato, Chiaki; Watanabe, Nobuhisa

2012-01-01

Structures of 3-isopropylmalate dehydrogenase were determined at pressures ranging from 0.1 to 650 MPa. Comparison of these structures gives a detailed picture of the swelling of a cavity at the dimer interface and the generation of a new cleft on the molecular surface, which are accompanied by water penetration. Hydrostatic pressure induces structural changes in proteins, including denaturation, the mechanism of which has been attributed to water penetration into the protein interior. In this study, structures of 3-isopropylmalate dehydrogenase (IPMDH) from Shewanella oneidensis MR-1 were determined at about 2 Å resolution under pressures ranging from 0.1 to 650 MPa using a diamond anvil cell (DAC). Although most of the protein cavities are monotonically compressed as the pressure increases, the volume of one particular cavity at the dimer interface increases at pressures over 340 MPa. In parallel with this volume increase, water penetration into the cavity could be observed at pressures over 410 MPa. In addition, the generation of a new cleft on the molecular surface accompanied by water penetration could also be observed at pressures over 580 MPa. These water-penetration phenomena are considered to be initial steps in the pressure-denaturation process of IPMDH

Adaptive Reference Control for Pressure Management in Water Networks

DEFF Research Database (Denmark)

Kallesøe, Carsten; Jensen, Tom Nørgaard; Wisniewski, Rafal

2015-01-01

Water scarcity is an increasing problem worldwide and at the same time a huge amount of water is lost through leakages in the distribution network. It is well known that improved pressure control can lower the leakage problems. In this work water networks with a single pressure actuator and several....... Subsequently, these relations are exploited in an adaptive reference control scheme for the actuator pressure that ensures constant pressure at the critical points. Numerical experiments underpin the results. © Copyright IEEE - All rights reserved....

Assessment of water pipes durability under pressure surge

Science.gov (United States)

Pham Ha, Hai; Minh, Lanh Pham Thi; Tang Van, Lam; Bulgakov, Boris; Bazhenova, Soafia

2017-10-01

Surge phenomenon occurs on the pipeline by the closing valve or pump suddenly lost power. Due to the complexity of the water hammer simulation, previous researches have only considered water hammer on the single pipe or calculation of some positions on water pipe network, it have not been analysis for all of pipe on the water distribution systems. Simulation of water hammer due to closing valve on water distribution system and the influence level of pressure surge is evaluated at the defects on pipe. Water hammer on water supply pipe network are simulated by Water HAMMER software academic version and the capacity of defects are calculated by SINTAP. SINTAP developed from Brite-Euram projects in Brussels-Belgium with the aim to develop a process for assessing the integrity of the structure for the European industry. Based on the principle of mechanical fault, indicating the size of defects in materials affect the load capacity of the product in the course of work, the process has proposed setting up the diagram to fatigue assessment defect (FAD). The methods are applied for water pipe networks of Lien Chieu district, Da Nang city, Viet Nam, the results show the affected area of wave pressure by closing the valve and thereby assess the greatest pressure surge effect to corroded pipe. The SINTAP standard and finite element mesh analysis at the defect during the occurrence of pressure surge which will accurately assess the bearing capacity of the old pipes. This is one of the bases to predict the leakage locations on the water distribution systems. Amount of water hammer when identified on the water supply networks are decreasing due to local losses at the nodes as well as the friction with pipe wall, so this paper adequately simulate water hammer phenomena applying for actual water distribution systems. The research verified that pipe wall with defect is damaged under the pressure surge value.

Nondestructive verification and assay systems for spent fuels

International Nuclear Information System (INIS)

Cobb, D.D.; Phillips, J.R.; Bosler, G.E.; Eccleston, G.W.; Halbig, J.K.; Hatcher, C.R.; Hsue, S.T.

1982-04-01

This is an interim report of a study concerning the potential application of nondestructive measurements on irradiated light-water-reactor (LWR) fuels at spent-fuel storage facilities. It describes nondestructive measurement techniques and instruments that can provide useful data for more effective in-plant nuclear materials management, better safeguards and criticality safety, and more efficient storage of spent LWR fuel. In particular, several nondestructive measurement devices are already available so that utilities can implement new fuel-management and storage technologies for better use of existing spent-fuel storage capacity. The design of an engineered prototype in-plant spent-fuel measurement system is approx. 80% complete. This system would support improved spent-fuel storage and also efficient fissile recovery if spent-fuel reprocessing becomes a reality

Swelling pressure and water absorption property of compacted granular bentonite during water absorption

International Nuclear Information System (INIS)

Oyamada, T.; Komine, H.; Murakami, S.; Sekiguchi, T.; Sekine, I.

2012-01-01

Document available in extended abstract form only. Bentonite is currently planned to be used as buffer materials in engineered barrier of radioactive waste disposal. Granular bentonites are expected as the materials used in constructions as buffer materials by in-situ compaction methods. After applying these buffer materials, it is expected that the condition of the buffer area changes in long-term by the seepage of groundwater into buffer area. Therefore, it is important to understand water movement and swelling behavior of the buffer materials for evaluating the performance of engineered barrier. In this study, we investigated water absorption property and swelling pressure of compacted granular bentonite. Specifically, the process of swelling pressure and amount of water absorption of granular bentonite-GX (Kunigel-GX, produced at the Tsukinuno mine in Japan) were observed by laboratory tests. To discuss the influence of maximum grain size of bentonite particle on swelling pressure and water absorption property, two types of samples were used. One is granular sample which is Bentonite-GX controlled under 2 mm the maximum grain size, the other is milled sample which is Bentonite-GX with the maximum grain size under 0.18 mm by milling with the agate mortar. In addition, the mechanism on the swelling pressure of compacted granular bentonite was considered and discussed. In the cases of granular sample, swelling pressure increases rapidly, then gradually continues to increase up to maximum value. In the cases of milled sample, swelling pressure also increases rapidly at first. However, then its value decreases before progressing of gradual increase continues. Especially, this trend was clearly observed at a relatively low dry density. At the peaks of these curves, the swelling pressure of granular samples is lower than that of milled samples. In addition, the increasing of swelling pressure by the time the peak observed during the process of swelling pressure from

Water Pressure Distribution on a Twin-Float Seaplane

Science.gov (United States)

Thompson, F L

1930-01-01

This is the second of a series of investigations to determine water pressure distribution on various types of seaplane floats and hulls, and was conducted on a twin-float seaplane. It consisted of measuring water pressures and accelerations on a TS-1 seaplane during numerous landing and taxiing maneuvers at various speeds and angles. The results show that water pressures as great as 10 lbs. per sq. in.may occur at the step in various maneuvers and that pressures of approximately the same magnitude occur at the stern and near the bow in hard pancake landings with the stern way down. At the other parts of the float the pressures are less and are usually zero or slightly negative for some distance abaft the step. A maximum negative pressure of 0.87 lb. Per square inch was measured immediately abaft the step. The maximum positive pressures have a duration of approximately one-twentieth to one-hundredth second at any given location and are distributed over a very limited area at any particular instant.

The velocity dependent dissolution of spent nuclear fuel in a geologic repository

International Nuclear Information System (INIS)

Nutt, W.M.

1990-02-01

A model describing the dissolution of fission products and transuranic isotopes from spent nuclear fuel into flowing ground water has been developed. This model is divided into two parts. The first part of the model calculates the temperature within a consolidated spent fuel waste form at a given time and ground water velocity. This model was used to investigate whether water flowing at rates representative of a geological repository located at Yucca Mountain, Nevada, will cool a wasteform consisting of consolidated spent nuclear fuel pins. Time and velocity dependent temperature profiles were generated. These profiles were input into the second model, which calculates the dissolution rate of waste isotopes from a spent fuel pin. Two dissolution limiting processes were modeled; the processes are dissolution limited by the solubility limit of an isotopes in the ground water, and dissolution limited by the diffusion of waste isotopes from the interior of a spent fuel pin to the surface where dissolution can occur

Pressurized-water coolant nuclear reactor steam generator

International Nuclear Information System (INIS)

Mayer, H.; Schroder, H.J.

1975-01-01

A description is given of a pressurized-water coolant nuclear reactor steam generator having a vertical housing for the steam generating water and containing an upstanding heat exchanger to which the pressurized-water coolant passes and which is radially surrounded by a guide jacket supporting a water separator on its top. By thermosiphon action the steam generating water flows upward through and around the heat exchanger within the guide chamber to the latter's top from which it flows radially outwardly and downwardly through a down draft space formed between the outside of the jacket and the housing. The water separator discharges separated water downwardly. The housing has a feedwater inlet opening adjacent to the lower portion of the heat exchanger, providing preheating of the introduced feedwater. This preheated feedwater is conveyed by a duct upwardly to a location where it mixes with the water discharged from the water separator

The influence of environmental conditions on lead transfer from spent gunshot to sediments and water: Other routes for Pb poisoning.

Science.gov (United States)

Binkowski, Łukasz J

2017-11-01

Lead (Pb) from spent gunshot and fishing sinkers is recognized as the main source of Pb poisoning among waterfowl. It is also suspected to directly pollute water and sediments, but no appropriate, comprehensive evaluation of this issue has so far been carried out. An experiment on Pb pellets in microcosms (n = 160) with two sediment types (mud and gravel), three water pH values (4, 7 and 9) and two wind levels (wind and windless simulation) was therefore run. Substantial differences in Pb transfer (measured with ICP-OES) between sediment types and pH levels of water were observed. Simulated wind conditions were a significant factor only for some variables and circumstances. The strongest Pb deposit to water and sediments occurred in mud microcosms with water of pH value of 4. Median pellet erosion during the experiment differed little between sediment types. The experiment revealed that Pb transfer from spent gunshot to the environment occurs only in specific environmental conditions. Copyright © 2017 Elsevier Ltd. All rights reserved.

Low cost sonoluminescence experiment in pressurized water

Energy Technology Data Exchange (ETDEWEB)

Bernal, L; Insabella, M [LADOP, University of Mar del Plata (Argentina); Bilbao, L [INFIP, University of Buenos Aires and CONICET (Argentina)

2012-06-19

We present a low cost design for demostration and mesurements of light emission from a sonoluminescence experiment. Using pressurized water introduced in an acrylic cylinder and one piezoelectric from an ultrasonic cleaner, we are able to generate cavitacion zones with emission of light. The use of argon to pressurize the water improves the emission an the light can be seen at naked eye in a softlit ambient.

Low cost sonoluminescence experiment in pressurized water

International Nuclear Information System (INIS)

Bernal, L; Insabella, M; Bilbao, L

2012-01-01

We present a low cost design for demostration and mesurements of light emission from a sonoluminescence experiment. Using pressurized water introduced in an acrylic cylinder and one piezoelectric from an ultrasonic cleaner, we are able to generate cavitacion zones with emission of light. The use of argon to pressurize the water improves the emission an the light can be seen at naked eye in a softlit ambient.

Heat insulation device for reactor pressure vessel in water

International Nuclear Information System (INIS)

Nakamura, Heiichiro; Tanaka, Yoshimi.

1993-01-01

Outer walls of a reactor pressure vessel are covered with water-tight walls made of metals. A heat insulation metal material is disposed between them. The water tight walls are joined by welding and flanges. A supply pipeline for filling gases and a discharge pipeline are in communication with the inside of the water tight walls. Further, a water detector is disposed in the midway of the gas discharge pipeline. With such a constitution, the following advantages can be attained. (1) Heat transfer from the reactor pressure vessel to water of a reactor container can be suppressed by filled gases and heat insulation metal material. (2) Since the pressure at the inside of the water tight walls can be equalized with the pressure of the inside of the reactor container, the thickness of the water-tight walls can be reduced. (3) Since intrusion of water to the inside of the walls due to rupture of the water tight walls is detected by the water detector, reactor scram can be conducted rapidly. (4) The sealing property of the flange joint portion is sufficient and detaching operation thereof is easy. (I.S.)

High pressure water jet mining machine

Science.gov (United States)

Barker, Clark R.

1981-05-05

A high pressure water jet mining machine for the longwall mining of coal is described. The machine is generally in the shape of a plowshare and is advanced in the direction in which the coal is cut. The machine has mounted thereon a plurality of nozzle modules each containing a high pressure water jet nozzle disposed to oscillate in a particular plane. The nozzle modules are oriented to cut in vertical and horizontal planes on the leading edge of the machine and the coal so cut is cleaved off by the wedge-shaped body.

Corrosion surveillance program of aluminum spent fuel elements in wet storage sites

International Nuclear Information System (INIS)

Linardi, E; Haddad, R

2012-01-01

Due to different degradation issues observed in aluminum-clad spent fuel during long term storage in water, the IAEA implemented in 1996 a Coordinated Research Project (CRP) and a Regional Project for Latin America, on Corrosion of Research Reactor Aluminum Clad Spent Fuel in Water. Argentine has been among the participant countries of these projects, carrying out spent fuel corrosion surveillance activities in its storage facilities. As a result of the research a large database on corrosion of aluminum-clad fuel has been generated. It was determined that the main types of corrosion affecting the spent fuel are pitting and galvanic corrosion due to contact with stainless steel. It was concluded that the quality of the water is the critical factor to control in a spent fuel storage facility. Another phase of the program is being conducted currently, which began in 2011 with the immersion of test racks in the RA1 reactor pool, and in the Research Reactor Spent Fuel Storage Facility (FACIRI), located in Ezeiza Atomic Center. This paper presents the results of the chemical analysis of the water performed so far, and its relationship with the examination of the coupons extracted from the sites (author)

Water Delivery--It's All about Pressure

Science.gov (United States)

Roman, Harry T.

2005-01-01

There is a great deal of wisdom in the old saying "water seeks its level." In fact, the concept has bearing on a very practical side of human life as well, since the public water delivery system is based on it. In this article, the author discusses the concept behind water pressure and describes how the water systems work based on this concept.…

Ultra-high pressure water jet: Baseline report

International Nuclear Information System (INIS)

1997-01-01

The ultra-high pressure waterjet technology was being evaluated at Florida International University (FIU) as a baseline technology. In conjunction with FIU's evaluation of efficiency and cost, this report covers the evaluation conducted for safety and health issues. It is a commercially available technology and has been used for various projects at locations throughout the country. The ultra-high pressure waterjet technology acts as a cutting tool for the removal of surface substrates. The Husky trademark pump feeds water to a lance that directs the high pressure water at the surface to be removed. The safety and health evaluation during the testing demonstration focused on two main areas of exposure. These were dust and noise. The dust exposure was found to be minimal, which would be expected due to the wet environment inherent in the technology, but noise exposure was at a significant level. Further testing for noise is recommended because of the outdoor environment where the testing demonstration took place. In addition, other areas of concern found were arm-hand vibration, ergonomics, heat stress, tripping hazards, electrical hazards, lockout/tagout, fall hazards, slipping hazards, hazards associated with the high pressure water, and hazards associated with air pressure systems

Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

International Nuclear Information System (INIS)

Chodak, P. III

1996-05-01

This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO 2 assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the 239 Pu and ≥90% total Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products

Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

Energy Technology Data Exchange (ETDEWEB)

Chodak, III, Paul [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

1996-05-01

This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO₂ assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the ²³⁹Pu and ≥90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

Spent fuel heatup following loss of water during storage

International Nuclear Information System (INIS)

Benjamin, A.S.; McCloskey, D.J.; Powers, D.A.; Dupree, S.A.

1979-03-01

An analysis of spent fuel heatup following a hypothetical accident involving drainage of the storage pool is presented. Computations based upon a new computer code called SFUEL have been performed to assess the effect of decay time, fuel element design, storage rack design, packing density, room ventilation, drainage level, and other variables on the heatup characteristics of the spent fuel and to predict the conditions under which clad failure will occur. Possible storage pool design modifications and/or onsite emergency action have also been considered

Electrochemical processing of spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Williamson, M. A.; Willit, J. L.; Barnes, L. A.; Figueroa, J.; Limmer, S. L.; Blaskovitz, R. [Argonne National Laboratory, Argonne (United States)

2008-08-15

Our work in developing the fuel cycles and electrochemical technologies needed for the treatment of spent light water reactor and spent fast reactor fuel is progressing well. Baseline flowsheets along with a theoretical material balance have been developed for treatment of each type of fuel. A discussion about the flowsheets provides the opportunity to present the status of our technology development activities and future research and development directions.

Electrochemical processing of spent nuclear fuel

International Nuclear Information System (INIS)

Williamson, M. A.; Willit, J. L.; Barnes, L. A.; Figueroa, J.; Limmer, S. L.; Blaskovitz, R.

2008-01-01

Our work in developing the fuel cycles and electrochemical technologies needed for the treatment of spent light water reactor and spent fast reactor fuel is progressing well. Baseline flowsheets along with a theoretical material balance have been developed for treatment of each type of fuel. A discussion about the flowsheets provides the opportunity to present the status of our technology development activities and future research and development directions

Spent Pot Lining Characterization Framework

Science.gov (United States)

Ospina, Gustavo; Hassan, Mohamed I.

2017-09-01

Spent pot lining (SPL) management represents a major concern for aluminum smelters. There are two key elements for spent pot lining management: recycling and safe storage. Spent pot lining waste can potentially have beneficial uses in co-firing in cement plants. Also, safe storage of SPL is of utmost importance. Gas generation of SPL reaction with water and ignition sensitivity must be studied. However, determining the feasibility of SPL co-firing and developing the required procedures for safe storage rely on determining experimentally all the necessary SPL properties along with the appropriate test methods, recognized by emissions standards and fire safety design codes. The applicable regulations and relevant SPL properties for this purpose are presented along with the corresponding test methods.

Performance Evaluation of Pressure Transducers for Water Impacts

Science.gov (United States)

Vassilakos, Gregory J.; Stegall, David E.; Treadway, Sean

2012-01-01

The Orion Multi-Purpose Crew Vehicle is being designed for water landings. In order to benchmark the ability of engineering tools to predict water landing loads, test programs are underway for scale model and full-scale water impacts. These test programs are predicated on the reliable measurement of impact pressure histories. Tests have been performed with a variety of pressure transducers from various manufacturers. Both piezoelectric and piezoresistive devices have been tested. Effects such as thermal shock, pinching of the transducer head, and flushness of the transducer mounting have been studied. Data acquisition issues such as sampling rate and anti-aliasing filtering also have been studied. The response of pressure transducers have been compared side-by-side on an impulse test rig and on a 20-inch diameter hemisphere dropped into a pool of water. The results have identified a range of viable configurations for pressure measurement dependent on the objectives of the test program.

Storage of spent fuel from light water reactors

International Nuclear Information System (INIS)

Wolkenhauer, W.C.

1976-01-01

The effects of possible inadequate nuclear fuel reprocessing capability upon a public utility, Washington Public Power Supply System, are studied. The possible alternatives for storing spent fuel are reviewed

Radiation characteristics of spent fuel of heavy-water research reactor during long-term storage

International Nuclear Information System (INIS)

Gerasimov, A.S.; Kiselev, G.V.; Myrtsymova, L.A.; Zaritskaya, T.S.

2002-01-01

Decay heat power and radiotoxicity by water of actinides and fission products from spent fuel of heavy-water research reactor RA were calculated for period of storage during 300000 years. Three variants of fuel enrichment by 235 U were considered: 2%, 21%, and 80%. The mass of 235 U in one fuel element was supposed to be the same for all variants of enrichment. The decay heat power of fission products in initial period is about 20 times higher than that of actinides. Decay heat power and radiotoxicity of actinides do not practically decrease during long period of time as they are determined by nuclides with very long half-life periods. (author)

Response of the water status of soybean to changes in soil water potentials controlled by the water pressure in microporous tubes

Science.gov (United States)

Steinberg, S. L.; Henninger, D. L.

1997-01-01

Water transport through a microporous tube-soil-plant system was investigated by measuring the response of soil and plant water status to step change reductions in the water pressure within the tubes. Soybeans were germinated and grown in a porous ceramic 'soil' at a porous tube water pressure of -0.5 kpa for 28 d. During this time, the soil matric potential was nearly in equilibrium with tube water pressure. Water pressure in the porous tubes was then reduced to either -1.0, -1.5 or -2.0 kPa. Sap flow rates, leaf conductance and soil, root and leaf water potentials were measured before and after this change. A reduction in porous tube water pressure from -0.5 to -1.0 or -1.5 kPa did not result in any significant change in soil or plant water status. A reduction in porous tube water pressure to -2.0 kPa resulted in significant reductions in sap flow, leaf conductance, and soil, root and leaf water potentials. Hydraulic conductance, calculated as the transpiration rate/delta psi between two points in the water transport pathway, was used to analyse water transport through the tube-soil-plant continuum. At porous tube water pressures of -0.5 to-1.5 kPa soil moisture was readily available and hydraulic conductance of the plant limited water transport. At -2.0 kPa, hydraulic conductance of the bulk soil was the dominant factor in water movement.

Reactivity and isotopic composition of spent PWR [pressurized-water-reactor] fuel as a function of initial enrichment, burnup, and cooling time

International Nuclear Information System (INIS)

Cerne, S.P.; Hermann, O.W.; Westfall, R.M.

1987-10-01

This study presents the reactivity loss of spent PWR fuel due to burnup in terms of the infinite lattice multiplications factor, k/sub ∞/. Calculations were performed using the SAS2 and CSAS1 control modules of the SCALE system. The k/sub ∞/ values calculated for all combinations of six enrichments, seven burnups, and five cooling times. The results are presented as a primary function of enrichment in both tabular and graphic form. An equation has been developed to estimate the tabulated values of k/sub ∞/'s by specifying enrichment, cooling time, and burnup. Atom densities for fresh fuel, and spent fuel at cooling times of 2, 10, and 20 years are included. 13 refs., 8 figs., 8 tabs

Nuclear spent fuel dry storage in the EWA reactor shaft

International Nuclear Information System (INIS)

Mieleszczenko, W.; Moldysz, A.; Hryczuk, A.; Matysiak, T.

2001-01-01

The EWA reactor was in operation from 1958 until February 1995. Then it was subjected to the decommissioning procedure. Resulting from a prolonged operation of Polish research reactors a substantial amount of nuclear spent fuel of various types, enrichment and degree of burnup have been accumulated. The technology of storage of spent nuclear fuel foresees the two stages of wet storing in a water pool (deferral period from tens to several dozens years) and dry storing (deferral period from 50 to 80 years). In our case the deferral time in the water environment is pretty significant (the oldest fuel elements have been stored in water for more than 40 years). Though the state of stored fuel elements is satisfactory, there is a real need for changing the storage conditions of spent fuel. The paper is covering the description of philosophy and conceptual design for construction of the spent fuel dry storage in the decommissioned EWA reactor shaft. (author)

Characteristics of hydride precipitation and reorientation in spent-fuel cladding

International Nuclear Information System (INIS)

Chung, H. M.; Strain, R. V.; Billone, M. C.

2000-01-01

The morphology, number density, orientation, distribution, and crystallographic aspects of Zr hydrides in Zircaloy fuel cladding play important roles in fuel performance during all phases before and after discharge from the reactor, i.e., during normal operation, transient and accident situations in the reactor, temporary storage in a dry cask, and permanent storage in a waste repository. In the past, partly because of experimental difficulties, hydriding behavior in irradiated fuel cladding has been investigated mostly by optical microscopy (OM). In the present study, fundamental metallurgical and crystallographic characteristics of hydride precipitation and reorientation were investigated on the microscopic level by combined techniques of OM and transmission electron and scanning electron microscopy (TEM and SEM) of spent-fuel claddings discharged from several boiling and pressurized water reactors (BWRs and PWRs). Defueled sections of standard and Zr-lined Zircaloy-2 fuel claddings, irradiated to fluences of ∼3.3 x 10 21 n cm -2 and ∼9.2 x 10 21 n cm -2 (E > 1 MeV), respectively, were obtained from spent fuel rods discharged from two BWRs. Sections of standard and low-tin Zircaloy-4 claddings, irradiated to fluences of ∼4.4 x 10 21 n cm -2 , ∼5.9 x 10 21 n cm -2 , and ∼9.6 x 10 21 n cm -2 (E > 1 MeV) in three PWRs, were also obtained. Microstructural characteristics of hydrides were analyzed in as-irradiated condition and after gas-pressurization-burst or expanding-mandrel tests at 292-325 C in Ar for some of the spent-fuel claddings. Analyses were also conducted of hydride habit plane, morphology, and reorientation characteristics on unirradiated Zircaloy-4 cladding that contained dense radial hydrides. Reoriented hydrides in the slowly cooled unirradiated cladding were produced by expanding-mandrel loading

Innovative leaching of cobalt and lithium from spent lithium-ion batteries and simultaneous dechlorination of polyvinyl chloride in subcritical water

Energy Technology Data Exchange (ETDEWEB)

Liu, Kang; Zhang, Fu-Shen, E-mail: fszhang@rcees.ac.cn

2016-10-05

Highlights: • A co-treatment process for recovery of Co and Li and simultaneous detoxification of PVC in subcritical water was proposed. • PVC was used as a hydrochloric acid source. • More than 95% Co and nearly 98% Li were leached under the optimum conditions. • Neither corrosive acid nor reducing agent was used. • The co-treatment process has technical, economic and environmental benefits over the traditional recovery processes. - Abstract: In this work, an effective and environmentally friendly process for the recovery of cobalt (Co) and lithium (Li) from spent lithium-ion batteries (LIBs) and simultaneously detoxification of polyvinyl chloride (PVC) in subcritical water was developed. Lithium cobalt oxide (LiCoO{sub 2}) power from spent LIBs and PVC were co-treated by subcritical water oxidation, in which PVC served as a hydrochloric acid source to promote metal leaching. The dechlorination of PVC and metal leaching was achieved simultaneously under subcritical water oxidation. More than 95% Co and nearly 98% Li were recovered under the optimum conditions: temperature 350 °C, PVC/LiCoO{sub 2} ratio 3:1, time 30 min, and a solid/liquid ratio 16:1 (g/L), respectively. Moreover, PVC was completely dechlorinated at temperatures above 350 °C without any release of toxic chlorinated organic compounds. Assessment on economical and environmental impacts revealed that the PVC and LiCoO{sub 2} subcritical co-treatment process had significant technical, economic and environmental benefits over the traditional hydrometallurgy and pyrometallurgy processes. This innovative co-treatment process is efficient, environmentally friendly and adequate for Co and Li recovery from spent LIBs and simultaneous dechlorination of PVC in subcritical water.

Worldwide spent fuel transportation logistics

International Nuclear Information System (INIS)

Best, R.E.; Garrison, R.F.

1978-01-01

This paper presents an overview of the worldwide transportation requirements for spent fuel. Included are estimates of numbers and types of shipments by mode and cask type for 1985 and the year 2000. In addition, projected capital and transportation costs are presented. For the year 1977 and prior years inclusive, there is a cumulative worldwide requirement for approximately 300 MTU of spent fuel storage at away-from-reactor (AFR) facilities. The cumulative requirements for years through 1985 are projected to be nearly 10,000 MTU, and for the years through 2000 the requirements are conservatively expected to exceed 60,000 MTU. These AFR requirements may be related directly to spent fuel transportation requirements. In total nearly 77,000 total cask shipments of spent fuel will be required between 1977 and 2000. These shipments will include truck, rail, and intermodal moves with many ocean and coastal water shipments. A limited number of shipments by air may also occur. The US fraction of these is expected to include 39,000 truck shipments and 14,000 rail shipments. European shipments to regional facilities are expected to be primarily by rail or water mode and are projected to account for 16,000 moves. Pacific basin shipments will account for 4500 moves. The remaining are from other regions. Over 400 casks will be needed to meet the transportation demands. Capital investment is expected to reach $800,000,000 in 1977 dollars. Cumulative transport costs will be a staggering $4.4 billion dollars

Determination of total plutonium content in spent nuclear fuel assemblies with the differential die-away self-interrogation instrument

Energy Technology Data Exchange (ETDEWEB)

Kaplan, Alexis C. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87544 (United States); Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 500 S State St., Ann Arbor, MI 48109 (United States); Henzl, Vladimir; Menlove, Howard O.; Swinhoe, Martyn T.; Belian, Anthony P. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87544 (United States); Flaska, Marek; Pozzi, Sara A. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 500 S State St., Ann Arbor, MI 48109 (United States)

2014-11-11

As a part of the Next Generation Safeguards Initiative Spent Fuel project, we simulate the response of the Differential Die-away Self-Interrogation (DDSI) instrument to determine total elemental plutonium content in an assayed spent nuclear fuel assembly (SFA). We apply recently developed concepts that relate total plutonium mass with SFA multiplication and passive neutron count rate. In this work, the multiplication of the SFA is determined from the die-away time in the early time domain of the Rossi-Alpha distributions measured directly by the DDSI instrument. We utilize MCNP to test the method against 44 pressurized water reactor SFAs from a simulated spent fuel library with a wide dynamic range of characteristic parameters such as initial enrichment, burnup, and cooling time. Under ideal conditions, discounting possible errors of a real world measurement, a root mean square agreement between true and determined total Pu mass of 2.1% is achieved.

A safety study on the wet storage of spent fuel

International Nuclear Information System (INIS)

Chun, Kwan Sik; Whang, Joo Ho; Lee, Hoo Kun; Choi, Jong Won; Lee, Jong Geun

1989-02-01

This study is to provide data related with a basic design of the spent fuel storage facility in the field of radiation and to establish the safety assessment methodology of away from reactor spent fuel storage facility. This is in progress and continue upto the year of 1991. The mathematical model which predict the quantity of environmental release of fission and corrosion products from spent fuel received and stored in wet storage facility operated in normal conditions was prepared. The decay characteristic of domestic spent fuels are analysed and then the coefficients for the prediction of the decay heat by simple formular was determined. This correlations could predict decay heat of spent fuel with ±10% difference from ORIGEN2 results. The release factor of cobalt out of PWR spent fuel in PIE pool is 7.97 x 10-12∼8.49 x 10-11 Ci/ sec-rod, which appears to be linear without being connected with the types of fuel defects, but that of cesium varies with the defect type and the exposure time in water. In water condition, release factor of uranium out of CANDU fuel pellets appears to be about 5 x 10-8/day, whose tendency is similar to that of cesium of the latter half of the exposure time of water. (Author)

Coolant mixing in pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Hoehne, T; Grunwald, G

1998-10-01

The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

Sensitivity of total stress to changes in externally applied water pressure in KBS-3 buffer bentonite

Energy Technology Data Exchange (ETDEWEB)

Harrington, J.F.; Birchall, D.J. [British Geological Survey, Chemical and Biological Hazards Programme, Kingsley Dunham Centre (United Kingdom)

2007-04-15

In the current Swedish repository design concept, composite copper and steel canisters containing spent nuclear fuel will be placed in large diameter disposal boreholes drilled into the floor of the repository tunnels. The space around each canister will be filled with pre-compacted bentonite which over time will draw in the surrounding ground water and swell, closing up any construction joints. However, for the purposes of performance assessment, it is necessary to consider the effect of glacial loading of a future repository and its impact on the mechanical behaviour of the bentonite, in particular, the sensitivity of total stress to changes in porewater pressure (backpressure). Two experimental histories have been undertaken using a custom-designed constant volume and radial flow (CVRF) apparatus. In both tests backpressure was varied in a number of incremental and decremental cycles while total stress, porewater pressure and volumetric flow rate were continuously monitored. The swelling pressure of the buffer clay at dry densities of 1.8 Mg/m{sup 3} and 1.61 Mg/m{sup 3} was determined to be around 5.5 MPa and 7.2 MPa respectively. For initial ascending porewater pressure histories the average proportionality factor {alpha} ranged from 0.86 and 0.92. Data exhibited a general trend of increasing {alpha} with increasing backpressure. In test Mx80-11 this was supported by analysis of the water inflow data which indicated a reduction in system compressibility. Asymptotic values of porewater pressure within the clay are in good agreement with externally applied backpressure values. Inspection of data provides no evidence for the development of hydraulic thresholds within the clay, subject to the boundary conditions of this test geometry. Analysis of the stress data demonstrates significant hysteresis between ascending and descending porewater pressure histories. The amount of hysteresis appears to be linked to the magnitude of the backpressure applied to the specimen

Sensitivity of total stress to changes in externally applied water pressure in KBS-3 buffer bentonite

International Nuclear Information System (INIS)

Harrington, J.F.; Birchall, D.J.

2007-04-01

In the current Swedish repository design concept, composite copper and steel canisters containing spent nuclear fuel will be placed in large diameter disposal boreholes drilled into the floor of the repository tunnels. The space around each canister will be filled with pre-compacted bentonite which over time will draw in the surrounding ground water and swell, closing up any construction joints. However, for the purposes of performance assessment, it is necessary to consider the effect of glacial loading of a future repository and its impact on the mechanical behaviour of the bentonite, in particular, the sensitivity of total stress to changes in porewater pressure (backpressure). Two experimental histories have been undertaken using a custom-designed constant volume and radial flow (CVRF) apparatus. In both tests backpressure was varied in a number of incremental and decremental cycles while total stress, porewater pressure and volumetric flow rate were continuously monitored. The swelling pressure of the buffer clay at dry densities of 1.8 Mg/m 3 and 1.61 Mg/m 3 was determined to be around 5.5 MPa and 7.2 MPa respectively. For initial ascending porewater pressure histories the average proportionality factor α ranged from 0.86 and 0.92. Data exhibited a general trend of increasing α with increasing backpressure. In test Mx80-11 this was supported by analysis of the water inflow data which indicated a reduction in system compressibility. Asymptotic values of porewater pressure within the clay are in good agreement with externally applied backpressure values. Inspection of data provides no evidence for the development of hydraulic thresholds within the clay, subject to the boundary conditions of this test geometry. Analysis of the stress data demonstrates significant hysteresis between ascending and descending porewater pressure histories. The amount of hysteresis appears to be linked to the magnitude of the backpressure applied to the specimen, suggesting some

Thermo-hydraulic behavior of saturated steam-water mixture in pressure vessel during injection of cold water

International Nuclear Information System (INIS)

Aya, Izuo; Kobayashi, Michiyuki; Inasaka, Fujio; Nariai, Hideki.

1983-01-01

The thermo-hydraulic behavior of saturated steam water mixture in a pressure vessel during injection of cold water was experimentally investigated with the Facility for Mixing Effect of Emergency Core Cooling Water. The dimensions of the pressure vessel used in the experiments were 284mm ID and 1,971mm height. 11 experiments were conducted without blowdown in order to comprehend the basic process excluding the effect of blowdown at injection of cold water. The initial pressure and water level, the injection flow rate and the size of injection nozzle were chosen as experimental parameters. Temperatures and void fractions at 6 elevations as well as pressure in the pressure vessel were measured, and new data especially on the pressure undershoot just after the initation of water injection and the vertical distribution of temperature and void fraction were gotten. The transients of pressure, average temperature and void fraction were caluculated using single-volume analysis code BLODAC-1V which is based on thermal equilibrium and so-called bubble gradient model. Some input parameters included in the analysis code were evaluated through the comparison of analysis with experimental data. Moreover, the observed pressure undershoot which is evaluated to be induced by a time lag of vapourization in water due to thermal nonequilibrium, was also discussed with the aid of another simple analysis model. (author)

High-pressure water electrolysis: Electrochemical mitigation of product gas crossover

International Nuclear Information System (INIS)

Schalenbach, Maximilian; Stolten, Detlef

2015-01-01

Highlights: • New technique to reduce gas crossover during water electrolysis • Increase of the efficiency of pressurized water electrolysis • Prevention of safety hazards due to explosive gas mixtures caused by crossover • Experimental realization for a polymer electrolyte membrane electrolyzer • Discussion of electrochemical crossover mitigation for alkaline water electrolysis - Abstract: Hydrogen produced by water electrolysis can be used as an energy carrier storing electricity generated from renewables. During water electrolysis hydrogen can be evolved under pressure at isothermal conditions, enabling highly efficient compression. However, the permeation of hydrogen through the electrolyte increases with operating pressure and leads to efficiency loss and safety hazards. In this study, we report on an innovative concept, where the hydrogen crossover is electrochemically mitigated by an additional electrode between the anode and the cathode of the electrolysis cell. Experimentally, the technique was applied to a proton exchange membrane water electrolyzer operated at a hydrogen pressure that was fifty times larger than the oxygen pressure. Therewith, the hydrogen crossover was reduced and the current efficiency during partial load operation was increased. The concept is also discussed for water electrolysis that is operated at balanced pressures, where the crossover of hydrogen and oxygen is mitigated using two additional electrodes

Experimental Study and Engineering Practice of Pressured Water Coupling Blasting

Directory of Open Access Journals (Sweden)

J. X. Yang

2017-01-01

Full Text Available Overburden strata movement in large space stope is the major reason that induces the appearance of strong mining pressure. Presplitting blasting for hard coal rocks is crucial for the prevention and control of strong pressure in stope. In this study, pressured water coupling blasting technique was proposed. The process and effect of blasting were analyzed by orthogonal test and field practice. Results showed that the presence of pressure-bearing water and explosive cartridges in the drill are the main influence factors of the blasting effect of cement test block. The high load-transmitting performance of pore water and energy accumulation in explosive cartridges were analyzed. Noxious substances produced during the blasting process were properly controlled because of the moistening, cooling, and diluting effect of pore water. Not only the goal of safe and static rock fragmentation by high-explosive detonation but also a combination of superdynamic blast loading and static loading effect of the pressured water was achieved. Then the practice of blasting control of hard coal rocks in Datong coal mine was analyzed to determine reasonable parameters of pressured water coupling blasting. A good presplitting blasting control effect was achieved for the hard coal rocks.

Transportation of spent nuclear fuels

International Nuclear Information System (INIS)

Meguro, Toshiichi

1976-01-01

The spent nuclear fuel taken out of reactors is cooled in the cooling pool in each power station for a definite time, then transported to a reprocessing plant. At present, there is no reprocessing plant in Japan, therefore the spent nuclear fuel is shipped abroad. In this paper, the experiences and the present situation in Japan are described on the transport of the spent nuclear fuel from light water reactors, centering around the works in Tsuruga Power Station, Japan Atomic Power Co. The spent nuclear fuel in Tsuruga Power Station was first transported in Apr. 1973, and since then, about 36 tons were shipped to Britain by 5 times of transport. The reprocessing plant in Japan is expected to start operation in Apr. 1977, accordingly the spent nuclear fuel used for the trial will be transported in Japan in the latter half of this year. Among the permission and approval required for the transport of spent nuclear fuel, the acquisition of the certificate for transport casks and the approval of land and sea transports are main tasks. The relevant laws are the law concerning the regulations of nuclear raw material, nuclear fuel and reactors and the law concerning the safety of ships. The casks used in Tsuruga Power Station and EXL III type, and the charging of spent nuclear fuel, the decontamination of the casks, the leak test, land transport with a self-running vehicle, loading on board an exclusive carrier and sea transport are briefly explained. The casks and the ship for domestic transport are being prepared. (Kato, I.)

Pressure dependence of dynamical heterogeneity in water

International Nuclear Information System (INIS)

Teboul, Victor

2008-01-01

Using molecular dynamics simulations we investigate the effect of pressure on the dynamical heterogeneity in water. We show that the effect of a pressure variation in water is qualitatively different from the effect of a temperature variation on the dynamical heterogeneity in the liquid. We observe a strong decrease of the aggregation of molecules of low mobility together with a decrease of the characteristic time associated with this aggregation. However, the aggregation of the most mobile molecules and the characteristic time of this aggregation are only slightly affected. In accordance with this result, the non-Gaussian parameter shows an important decrease with pressure while the characteristic time t* of the non-Gaussian parameter is only slightly affected. These results highlight then the importance of pressure variation investigations in low temperature liquids on approach to the glass transition

The phase diagram of water at negative pressures: virtual ices.

Science.gov (United States)

Conde, M M; Vega, C; Tribello, G A; Slater, B

2009-07-21

The phase diagram of water at negative pressures as obtained from computer simulations for two models of water, TIP4P/2005 and TIP5P is presented. Several solid structures with lower densities than ice Ih, so-called virtual ices, were considered as possible candidates to occupy the negative pressure region of the phase diagram of water. In particular the empty hydrate structures sI, sII, and sH and another, recently proposed, low-density ice structure. The relative stabilities of these structures at 0 K was determined using empirical water potentials and density functional theory calculations. By performing free energy calculations and Gibbs-Duhem integration the phase diagram of TIP4P/2005 was determined at negative pressures. The empty hydrates sII and sH appear to be the stable solid phases of water at negative pressures. The phase boundary between ice Ih and sII clathrate occurs at moderate negative pressures, while at large negative pressures sH becomes the most stable phase. This behavior is in reasonable agreement with what is observed in density functional theory calculations.

Evaluation of the Compressive Strength of Cement-Spent Resins Matrix Mixed with Bio char

International Nuclear Information System (INIS)

Zalina Laili; Muhamad Samudi Yasir; Zalina Laili; Mohd Abdul Wahab; Nur Azna Mahmud; Nurfazlina Zainal Abidin

2015-01-01

The evaluation of compressive strength of cement-spent resins matrix mixed with bio char was investigated. In this study, bio char with different percentage (5 %, 8 %, 11 % 14 % and 18 %) was used as alternative admixture material for cement solidification of spent resins. Some properties of the physical and chemical of spent resins and bio char were also investigated. The performance of cemented spent resins with the addition of bio char was evaluated based on their compressive strength and the water resistance test. The compressive strength was evaluated at three different curing periods of 7, 14 and 28 days, while 4 weeks of immersion in distilled water was chosen for water resistance test. The result indicated that the compressive strength at 7, 14 and 28 days of curing periods were above the minimum criterion for example > 3.45 MPa of acceptable level for cemented waste form. Statistical analysis showed that there was no significant relationship between the compressive strength of the specimen and the percentage of bio char content. Result from the water resistance test showed that only one specimen that contained of 5 % of bio char failed the water resistance test due to the high of spent resins/ bio char ratio. The compressive strength of cement solidified spent resins was found increased after the water resistance test indicating further hydration occurred after immersed in water. The results of this study also suggest that the specimen with 8 %, 11 %, 14 % and 18 % of bio char content were resistance in water and suitable for the leaching study of radionuclides from cement-bio char-spent resins matrix. (author)

Characteristics of spent nuclear fuel

International Nuclear Information System (INIS)

Notz, K.J.

1988-04-01

The Office of Civilian Radioactive Waste Management (OCRWM) is responsible for the spent fuels and other wastes that will, or may, eventually be disposed of in a geological repository. The two major sources of these materials are commercial light-water reactor (LWR) spent fuel and immobilized high-level waste (HLW). Other wastes that may require long-term isolation include non-LWR spent fuels and miscellaneous sources such as activated metals. This report deals with spent fuels, but for completeness, the other sources are described briefly. Detailed characterizations are required for all of these potential repository wastes. These characteristics include physical, chemical, and radiological properties. The latter must take into account decay as a function of time. In addition, the present inventories and projected quantities of the various wastes are needed. This information has been assembled in a Characteristics Data Base which provides data in four formats: hard copy standard reports, menu-driven personal computer (PC) data bases, program-level PC data bases, and mainframe computer files. 5 refs., 3 figs., 4 tabs

Pressure: the politechnics of water supply in Mumbai.

Science.gov (United States)

Anand, Nikhil

2011-01-01

In Mumbai, most all residents are delivered their daily supply of water for a few hours every day, on a water supply schedule. Subject to a more precarious supply than the city's upper-class residents, the city's settlers have to consistently demand that their water come on “time” and with “pressure.” Taking pressure seriously as both a social and natural force, in this article I focus on the ways in which settlers mobilize the pressures of politics, pumps, and pipes to get water. I show how these practices not only allow settlers to live in the city, but also produce what I call hydraulic citizenship—a form of belonging to the city made by effective political and technical connections to the city's infrastructure. Yet, not all settlers are able to get water from the city water department. The outcomes of settlers' efforts to access water depend on a complex matrix of socionatural relations that settlers make with city engineers and their hydraulic infrastructure. I show how these arrangements describe and produce the cultural politics of water in Mumbai. By focusing on the ways in which residents in a predominantly Muslim settlement draw water despite the state's neglect, I conclude by pointing to the indeterminacy of water, and the ways in which its seepage and leakage make different kinds of politics and publics possible in the city.

Options and processes for spent catalyst handling and utilization.

Science.gov (United States)

Marafi, M; Stanislaus, A

2003-07-18

The quantity of spent hydroprocessing catalysts discarded as solid wastes in the petroleum refining industries has increased remarkably in recent years due to a rapid growth in the hydroprocessing capacity to meet the rising demand for low-sulfur fuels. Due to their toxic nature, spent hydroprocessing catalysts have been branded as hazardous wastes, and the refiners are experiencing pressure from environmental authorities to handle them safely. Several alternative methods such as reclamation of metals, rejuvenation and reuse, disposal in landfills and preparation of useful materials using spent catalysts as raw materials are available to deal with the spent catalyst problem. The technical feasibility as well as the environmental and economic aspects of these options are reviewed. In addition, details of two bench-scale processes, one for rejuvenation of spent hydroprocessing catalysts, and the other for producing non-leachable synthetic aggregate materials that were developed in this laboratory, are presented in this paper.

Results from Nevada Nuclear Waste Storage Investigations (NNWSI) Series 3 spent fuel dissolution tests

International Nuclear Information System (INIS)

Wilson, C.N.

1990-06-01

The dissolution and radionuclide release behavior of spent fuel in groundwater is being studied by the Yucca Mountain Project (YMP), formerly the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. Specimens prepared from pressurized water reactor fuel rod segments were tested in sealed stainless steel vessels in Nevada Test Site J-13 well water at 85 degree C and 25 degree C. The test matrix included three specimens of bare-fuel particles plus cladding hulls, two fuel rod segments with artificially defected cladding and water-tight end fittings, and an undefected fuel rod section with watertight end fittings. Periodic solution samples were taken during test cycles with the sample volumes replenished with fresh J-13 water. Test cycles were periodically terminated and the specimens restarted in fresh J-13 water. The specimens were run for three cycles for a total test duration of 15 months. 22 refs., 32 figs., 26 tabs

Water hammer characteristics of integral pressurized water reactor primary loop

International Nuclear Information System (INIS)

Zuo, Qiaolin; Qiu, Suizheng; Lu, Wei; Tian, Wenxi; Su, Guanghui; Xiao, Zejun

2013-01-01

Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions

Water hammer characteristics of integral pressurized water reactor primary loop

Energy Technology Data Exchange (ETDEWEB)

Zuo, Qiaolin [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Qiu, Suizheng, E-mail: szqiu@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Lu, Wei; Tian, Wenxi; Su, Guanghui [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Xiao, Zejun [Nuclear Power Institute of China, Chengdu, Sichuan 610041 (China)

2013-08-15

Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions.

Spent fuel from RA reactor inspection of state and options for management

International Nuclear Information System (INIS)

Aden, V.G.; Bulkin, S. Yu.; Sokolov, A. V.; Matausek, M.V.; Vukadin, Z.

2001-01-01

About five thousand spent fuel elements from RA reactor have been stored for over 30 years in sealed aluminum barrels in the spent fuel storage pool. This way of storage does not provide complete information about the state of spent fuel elements or the medium inside the barrels, like pressure or radioactivity. The technology has been developed and the equipment has been manufactured to inspect the state of the spent fuel and to reduce eventual internal pressure inside the aluminum barrels. The realization of this technology was started in 1999 but due to political and financial difficulties was not completed. In September the year 2000 the work was restarted. Two different ways of RA reactor spent fuel elements preparation for transportation or long-term storage are considered: 'all fuel elements canning without leak-tightness testing' and 'all fuel elements leak-tightness testing'. It is believed that the first option offers several distinct advantages, which can be summarized as: greater reliability in the course of transportation or dry storage. Higher safety for workers. Lower expenditures for non-standard equipment manufacturing. Shorter duration of work. (author)

Pressurized water reactor fuel rod design methodology

International Nuclear Information System (INIS)

Silva, A.T.; Esteves, A.M.

1988-08-01

The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

In Situ Raman Study of Liquid Water at High Pressure.

Science.gov (United States)

Romanenko, Alexandr V; Rashchenko, Sergey V; Goryainov, Sergey V; Likhacheva, Anna Yu; Korsakov, Andrey V

2018-06-01

A pressure shift of Raman band of liquid water (H 2 O) may be an important tool for measuring residual pressures in mineral inclusions, in situ barometry in high-pressure cells, and as an indicator of pressure-induced structural transitions in H 2 O. However, there was no consensus as to how the broad and asymmetric water Raman band should be quantitatively described, which has led to fundamental inconsistencies between reported data. In order to overcome this issue, we measured Raman spectra of H 2 O in situ up to 1.2 GPa using a diamond anvil cell, and use them to test different approaches proposed for the description of the water Raman band. We found that the most physically meaningful description of water Raman band is the decomposition into a linear background and three Gaussian components, associated with differently H-bonded H 2 O molecules. Two of these components demonstrate a pronounced anomaly in pressure shift near 0.4 GPa, supporting ideas of structural transition in H 2 O at this pressure. The most convenient approach for pressure calibration is the use of "a linear background + one Gaussian" decomposition (the pressure can be measured using the formula P (GPa) = -0.0317(3)·Δν G (cm -1 ), where Δν G represents the difference between the position of water Raman band, fitted as a single Gaussian, in measured spectrum and spectrum at ambient pressure).

A review on water fault diagnosis of PEMFC associated with the pressure drop

International Nuclear Information System (INIS)

Pei, Pucheng; Li, Yuehua; Xu, Huachi; Wu, Ziyao

2016-01-01

Highlights: • Reviewed the effect factors and estimations of pressure drop associated with water fault diagnosis. • Reviewed pressure drop-based water fault diagnosis using different indicators. • Deviation of pressure drop is used frequently to diagnose water fault. • Reviewed recovery strategies based on pressure drop used in commercial PEMFC. • Merits, demerits and application prospects of pressure drop-based water fault diagnosis are discussed. - Abstract: The pressure difference between the inlet and outlet of the reactant in fuel cells is called the pressure drop, which is related to the water amount inside the fuel cells. In recent years there have been many studies that used the pressure drop to detect the water content and diagnose water fault of proton exchange membrane fuel cells (PEMFCs). To our knowledge, there has not been a systematic review of these studies. In this paper, the effect variables of pressure drop are reviewed firstly. Then estimations of the theoretical pressure drop are reviewed mainly based on the following four aspects: Bernoulli’s equation, two-phase flow multiplier, Darcy’s law and artificial intelligence. Afterward, the water fault diagnosis based on the pressure drop using the following six indicators are reviewed: indicator of direct pressure drop, its deviation, frequency, multiplier, the ratio of pressure drop to flow rate and the flooding degree. In addition, the strategies of water fault recovery are also summarized. Finally the merits, demerits and application prospects of pressure drop-based water fault diagnosis are presented.

Thermal Analysis Evaluation of Spent Fuel Storage Rack for Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Sangjin; Oh, Jinho; Kwak, Jinsung; Lee, Jongmin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Spent fuel storage rack is to store spent fuel assemblies. The spent fuel storage rack is submerged in the designated pool for cooling. Due to the condition change of the pool water, the effect of thermal load on spent fuel storage rack must be analyzed and evaluated. In this paper, thermal stress analysis is performed and evaluated on a spent fuel storage rack. For thermal stress evaluation of the spent fuel storage rack, load combinations and allowable criteria in ASME Sec. III NB-3220 are applied. In cases of A-1 and B-1, the same temperature applied on the whole model, thermal stress doesn't occur because there is no constraint about the thermal expansion. The support frame is located on the pool bottom in free standing type and the racks are located in the support frame with enough space. Thermal expansion was considered and reflected in the design of spent fuel storage rack in advance. Thermal stress analysis is performed and evaluated on a spent fuel storage rack with consideration of pool water temperature variation. The thermal analysis including a linear heat transfer and the thermal stress analysis is performed for the racks and support frame and resulted stresses are within allowable criteria.

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

SIMULATION OF NEGATIVE PRESSURE WAVE PROPAGATION IN WATER PIPE NETWORK

Directory of Open Access Journals (Sweden)

Tang Van Lam

2017-11-01

Full Text Available Subject: factors such as pipe wall roughness, mechanical properties of pipe materials, physical properties of water affect the pressure surge in the water supply pipes. These factors make it difficult to analyze the transient problem of pressure evolution using simple programming language, especially in the studies that consider only the magnitude of the positive pressure surge with the negative pressure phase being neglected. Research objectives: determine the magnitude of the negative pressure in the pipes on the experimental model. The propagation distance of the negative pressure wave will be simulated by the valve closure scenarios with the help of the HAMMER software and it is compared with an experimental model to verify the quality the results. Materials and methods: academic version of the Bentley HAMMER software is used to simulate the pressure surge wave propagation due to closure of the valve in water supply pipe network. The method of characteristics is used to solve the governing equations of transient process of pressure change in the pipeline. This method is implemented in the HAMMER software to calculate the pressure surge value in the pipes. Results: the method has been applied for water pipe networks of experimental model, the results show the affected area of negative pressure wave from valve closure and thereby we assess the largest negative pressure that may appear in water supply pipes. Conclusions: the experiment simulates the water pipe network with a consumption node for various valve closure scenarios to determine possibility of appearance of maximum negative pressure value in the pipes. Determination of these values in real-life network is relatively costly and time-consuming but nevertheless necessary for identification of the risk of pipe failure, and therefore, this paper proposes using the simulation model by the HAMMER software. Initial calibration of the model combined with the software simulation results and

Neutron fluence determination for light water reactor pressure vessels

International Nuclear Information System (INIS)

Gold, R.

1994-01-01

A general description of limitations that exist in pressure vessel neutron fluence determinations for commercial light water reactors is presented. Complexity factors that arise in light water reactor pressure vessel neutron fluence calculations are identified and used to analyze calculational limitations. Two broad categories of calculational limitations are introduced, namely benchmark field limitations and deep penetration limitations. Explicit examples of limitations that can arise in each of these two broad categories are presented. These limitations are used to show that the recent draft regulatory guide for the determination of pressure vessel neutron fluence, developed by the Nuclear Regulatory Commission, is based upon procedures and assumptions that are not valid. To eliminate the complexity and limitations of calculational methods, it is recommended that the determination of light water reactor pressure vessel neutron fluence be based upon experiment. Recommendations for improved methods of pressure vessel surveillance neutron dosimetry are advanced

Spent fuel management in Spain

International Nuclear Information System (INIS)

Gago, J.A.; Gravalos, J.M.

1996-01-01

There are presently nine Light Water Reactors in operation, representing around a 34% of the overall electricity production. In the early years, a small amount of spent fuel was sent to be reprocessed, although this policy was cancelled in favor of the open cycle option. A state owned company, ENRESA, was created in 1984, which was given the mandate to manage all kinds of radioactive wastes generated in the country. Under the present scenario, a rough overall amount of 7000 tU of spent fuel will be produced during the lifetime of the plants, which will go into final disposal. (author)

Transportation of spent fuel from light water reactors

International Nuclear Information System (INIS)

Bernard, H.

1993-01-01

The French 'Compagnie Generale des Matieres Nucleaires' - COGEMA - is involved in the whole nuclear fuel cycle about 20 years. Among the different parts of the cycle, the Transport of Radioactive Materials, acting as a link between the differents plants has a great importance. As nuclear material transportation is the only fuel cycle step to be performed on public grounds, the industrial task has to be performed with the utmost stringent safety criteria. COGEMA and associates is now operating a fully mature commercial activity, with some 300 spent fuel shipments per year from its reprocessing customer's reactors to the LA HAGUE plant, either by rail, road or sea. The paper will review the organization of COGEMA transportation business, the level of technology with an update of the casks used for spent fuel, and the operational experience, with a particular view of the maintenance policy. (author)

High-Pressure Liquid Chromatography of Irradiated Nuclear Fue - Separation of Neodymium for Burn-up Determination

DEFF Research Database (Denmark)

Larsen, N. R.

1979-01-01

Neodymium is separated from solutions of spent nuclear fuel by high-pressure liquid chromatography in methanol-nitric acid-water media using an anion-exchange column. Chromatograms obtained by monitoring at 280 nm, illustrate the difficulties especially with the fission product ruthenium in nuclear...

Water activities in Forsmark (Part II). The final disposal facility for spent fuel: water activities above ground; Vattenverksamhet i Forsmark (del II). Slutfoervarsanlaeggningen foer anvaent kaernbraensle: Vattenverksamheter ovan mark

Energy Technology Data Exchange (ETDEWEB)

Werner, Kent [EmpTec (Sweden); Hamren, Ulrika; Collinder, Per [Ekologigruppen AB (Sweden); Ridderstolpe, Peter [WRS Uppsala AB (Sweden)

2010-09-15

The construction of the repository for spent nuclear fuel in Forsmark is associated with a number of measures above ground that constitute water operations according to Chapter 11 in the Swedish Environmental Code. This report, which is an appendix to the Environmental Impact Assessment, describes these water operations, their effects and consequences, and planned measures

Is high-pressure water the cradle of life?

International Nuclear Information System (INIS)

Bassez, Marie-Paule

2003-01-01

Several theories have been proposed for the synthesis of prebiotic molecules. This letter shows that the structure of supercritical water, or high-pressure water, could trigger prebiotic synthesis and the origin of life deep in the oceans, in hydrothermal vent systems. Dimer geometries of high-pressure water may have a point of symmetry and a zero dipole moment. Consequently, simple apolar molecules found in submarine hydrothermal vent systems will dissolve in the apolar environment provided by the apolar form of the water dimer. Apolar water could be the medium which helps precursor molecules to concentrate and react more efficiently. The formation of prebiotic molecules could thus be linked to the structure of the water inside chimney nanochannels and cavities where hydrothermal piezochemistry and shock wave chemistry could occur. (letter to the editor)

Development of Integrity Evaluation Technology for the Long-term Spent Fuel Dry Storage System (1st year Report)

International Nuclear Information System (INIS)

Choi, Jong Won; Kook, Dong Hak; Kim, Jun Sub

2010-05-01

Korea has operated 16 Pressurized Water Reactors(PWR) and has a plan to construct additional nuclear power reactors as only PWR. This causes a big issue of PWR spent fuel accumulation problem now and in the future. KRMC(Korea Radioactive waste Management Coorporation) which was established in 2009 is charged with managing all kinds of radioactive waste that is produced in Korea. KRMC is considering spent fuel dry storage as an option to solve this spent fuel problem and developing the related engineering techniques. KAERI(Korea Atomic Energy Research Institute) also participated in this development and focused on evaluating the spent fuel dry storage system integrity for a long term operation. This report is the first year research product. The aims of the first year work scope are surveying and analyzing models which could anticipate degradation phenomena of the all dry storage components(spent fuel, structure materials, and equipment materials) and selecting items of the tests which are planned to perform in the next project stage. The major work areas consist of 'spent fuel degradation evaluation model development', 'test senario development', 'long-term evaluation of structural material characteristics', and 'dry storage system structure degradation model development'. These works were successfully achieved. This report is expected to contribute for the second year work which includes degradation model development and test senario development, and next project stage

Capillarity Induced Negative Pressure of Water Plugs in Nanochannels

NARCIS (Netherlands)

Tas, Niels Roelof; Mela, P.; Kramer, Tobias; Berenschot, Johan W.; van den Berg, Albert

2003-01-01

We have found evidence that water plugs in hydrophilic nanochannels can be at significant negative pressure due to tensile capillary forces. The negative pressure of water plugs in nanochannels induces bending of the thin channel capping layer, which results in a visible curvature of the liquid

Spent LWR fuel leach tests: Waste Isolation Safety Assessment program

International Nuclear Information System (INIS)

Katayama, Y.B.

1979-04-01

Spent light-water-reactor (LWR) fuels with burnups of 54.5, 28 and 9 MWd/kgU were leach-tested in deionized water at 25 0 C. Fuel burnup has no apparent effect on the calculated leach rates based upon the behavior of 137 Cs and 239+240 Pu. A leach test of 54.5 MWd/kgU spent fuel in synthetic sea brine showed that the cesium-based leach rate is lower in sea brine than in deionized water. A rise in the leach rate was observed after approximately 600 d of cumulative leaching. During the rise, the leach rate for all the measured radionuclides become nearly equal. Evidence suggests that exposure of new surfaces to the leachant may cause the increase. As a result, experimental work to study leaching mechanisms of spent fuel has been initiated. 22 figures

Heat transfer in a spent fuel pool concept containing PWR, Hybrid ADS-Fission, and VHTR spent fuels

Energy Technology Data Exchange (ETDEWEB)

Faria, Fernando P.; Cardoso, Fabiano; Salomé, Jean A.D.; Velasquez, Carlos E.; Pereira, Claubia, E-mail: fernandopereirabh@gmail.com, E-mail: fabinuclear@yahoo.com.br, E-mail: jadsalome@yahoo.com.br, E-mail: carlosvelcab@hotmail.com, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-07-01

Thermal evaluation under wet storage conditions of spent fuels (SF) of the types UO{sub 2} discharged from Pressurized Water Reactor (PWR) and Very High-temperature Reactor (VHTR), and (Th,TRU)O{sub 2} from Accelerator-Driven Subcritical Reactor System (ADS) and VHTR are presented. The analyzes are in the absence of an external cooling system of the pool, and the goal is to compare the water boiling time of the pool storing these different types of SF, at time t=0 year after reactor discharge. Two techniques were implemented. In the first one, all the materials of the fuel elements are considered. In the second, the SF is treated as holes inside the pool, assuming the heat transfer directly from the SF to the water. Results from first technique show that the boiling time (T{sub b}) ranged from 23 minutes for (Th,TRU)O{sub 2} from VHTR to 3 hours for UO{sub 2} from VHTR, while for the second technique, T{sub b} ranged from 10 minutes for (Th,TRU)O{sub 2} from VHTR to 2.7 hours for UO{sub 2} from VHTR. The discrepancies between Tb from both techniques reveal that the pathways considered for the heat transfer are crucial to the results. The thermal studies used the module CFX of the ANSYS Workbench 16.2 - student version. (author)

Spent fuel reprocessing options

International Nuclear Information System (INIS)

2008-08-01

The objective of this publication is to provide an update on the latest developments in nuclear reprocessing technologies in the light of new developments on the global nuclear scene. The background information on spent fuel reprocessing is provided in Section One. Substantial global growth of nuclear electricity generation is expected to occur during this century, in response to environmental issues and to assure the sustainability of the electrical energy supply in both industrial and less-developed countries. This growth carries with it an increasing responsibility to ensure that nuclear fuel cycle technologies are used only for peaceful purposes. In Section Two, an overview of the options for spent fuel reprocessing and their level of development are provided. A number of options exist for the treatment of spent fuel. Some, including those that avoid separation of a pure plutonium stream, are at an advanced level of technological maturity. These could be deployed in the next generation of industrial-scale reprocessing plants, while others (such as dry methods) are at a pilot scale, laboratory scale or conceptual stage of development. In Section Three, research and development in support of advanced reprocessing options is described. Next-generation spent fuel reprocessing plants are likely to be based on aqueous extraction processes that can be designed to a country specific set of spent fuel partitioning criteria for recycling of fissile materials to advanced light water reactors or fast spectrum reactors. The physical design of these plants must incorporate effective means for materials accountancy, safeguards and physical protection. Section four deals with issues and challenges related to spent fuel reprocessing. The spent fuel reprocessing options assessment of economics, proliferation resistance, and environmental impact are discussed. The importance of public acceptance for a reprocessing strategy is discussed. A review of modelling tools to support the

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.)

License considerations of the temporary storage in dry of the nuclear spent fuel of light water reactors; Consideraciones de licenciamiento del almacenamiento temporal en seco del combustible gastado nuclear de reactores de agua ligera

Energy Technology Data Exchange (ETDEWEB)

Bazan L, A.; Vargas A, A.; Cardenas J, J. B., E-mail: ariadna.bazan@cfe.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Carretera Cardel-Nautla Km 42.5, Alto Lucero, Veracruz (Mexico)

2011-11-15

The spent fuel of the nuclear power plants of light water is usually stored in cells or frames inside steel coating pools. The water of the spent fuel pool has a double function: it serves as shielding or barrier for the radiation that emits the spent fuel and, on the other hand, to cool it in accordance with its decay in the time. The administration policies of the spent fuel vary of some countries to other, resulting common to all the cases this initial stage of cooling in the pools, after its irradiation in the reactor. When is not possible to increase more this capacity, usually, technologies of temporary storage in dry of the spent fuel in independent facilities are used. The storage in dry of the spent fuel differs of the storage in the fuel pools making use of gas instead of water as coolant and using metal or concrete instead of the water like barrier against the radiation. The storage industry in dry offers a great variety of technologies, which should be certified by the respective nuclear regulator entity before its use. (Author)

Reactor vessel pressure transient protection for pressurized water reactors

International Nuclear Information System (INIS)

Zech, G.

1978-09-01

During the past few years the NRC has been studying the issue of protection of the reactor pressure vessels at Pressurized Water Reactors (PWRs) from transients when the vessels are at a relatively low temperature. This effort was prompted by concerns related to the safety margins available to vessel damage as a result of such events. Nuclear Reactor Regulation Category A Technical Activity No. A-26 was established to set forth the NRC plan for resolution of the generic aspects of this safety issue. The purpose of the report is to document the completion of this generic technical activity

Effects of actinide compositional variability in the U.S. spent fuel inventory on partitioning-transmutation systems

International Nuclear Information System (INIS)

Ludwig, S.B.; Michaels, G.E.; Hanson, B.D.

1993-01-01

The partitioning and transmutation concept (P-T) has as a mission the reduction by many orders of magnitude of certain undesirable nuclides in the waste streams. Given that only a very small fiction of spent fuel can be rejected by a P-T enterprise, a P-T system must therefore be capable of accommodating a wide range of spent fuel characteristics. Variability of nuclide composition (i.e. the feed material for transmutation devices) may be important because virtually all transmutation systems propose to configure TRU nuclides recovered from discharged LWR fuel in critical or near-critical cores. To date, all transmutation system core analyses assume nonvariable nuclide concentrations for startup and recycle cores. Using the Department of Energy (DOES) Characteristic Data Base (CDB) and the ORIGEN2 computer code, the current and projected spent fuel discharges until the year 2016 have been categorized according to combinations of fuel burnup, initial enrichment, fuel age (cooling time) and reactor type (boiling-water or pressurized-water reactor). In addition to quantifying the variability of nuclide composition in current and projected LWR fuel discharge, the variability of the infinite multiplication factor (K ∞ ) is calculated for both fast (ALMR) and thermal (accelerator-based) transmuter systems. It is shown that actinide compositional variations are potentially significant and warrant further investigation. (authors)

Test plan for Series 3 NNWSI spent fuel leaching/dissolution tests

International Nuclear Information System (INIS)

Wilson, C.N.

1986-04-01

The Series 3 tests will differ from the Series 2 tests in that the Series 3 tests will be run at 85 0 C (J-13 water) in sealed 304 stainless steel (SS) test vessels. The current NNWSI reference spent fuel container material is 304L SS. The candidate NNWSI repository horizon is above the water table, and 95 0 C (boiling temperature at the repository elevation) is the maximum liquid water temperature expected to contact spent fuel in the repository

Long-Term Dry Storage of High Burn-Up Spent Pressurized Water Reactor (PWR) Fuel in TAD (Transportation, Aging, and Disposal) Containers

International Nuclear Information System (INIS)

Hwang, Yong Soo

2008-12-01

A TAD canister, in conjunction with specially-designed over-packs can accomplish the functions of transportation, aging, and disposal (TAD) in the management of spent nuclear fuel (SNF). Industrial dry cask systems currently available for SNF are licensed for storage-only or for dual-purpose (i.e., storage and transportation). By extending the function to include the indefinite storage and perhaps, eventual geologic disposal, the TAD canister would have to be designed to enhance, among others, corrosion resistance, thermal stability, and criticality-safety control. This investigative paper introduces the use of these advanced iron-based, corrosion-resistant materials for SNF transportation, aging, and disposal.The objective of this investigative project is to explore the interest that KAERI would research and develop its specific SAM coating materials for the TAD canisters to satisfy the requirements of corrosion-resistance, thermal stability, and criticality-controls for long-term dry storage of high burn-up spent PWR fuel

Tensile Strength of Water Exposed to Pressure Pulses

DEFF Research Database (Denmark)

Andersen, Anders Peter; Mørch, Knud Aage

2012-01-01

at an extended water-solid interface by imposing a tensile stress pulse which easily causes cavitation. Next, a compressive pulse of duration ~1 ms and a peak intensity of a few bar is imposed prior to the tensile stress pulse. A dramatic increase of the tensile strength is observed immediately after......It is well known that pressurization for an extended period of time increases the tensile strength of water, but little information is available on the effect of pressure pulses of short duration. This is addressed in the present paper where we first measure the tensile strength of water...

Analytical study on water hammer pressure in pressurized conduits with a throttled surge chamber for slow closure

Directory of Open Access Journals (Sweden)

Yong-liang Zhang

2010-06-01

Full Text Available This paper presents an analytical investigation of water hammer in a hydraulic pressurized pipe system with a throttled surge chamber located at the junction between a tunnel and a penstock, and a valve positioned at the downstream end of the penstock. Analytical formulas of maximum water hammer pressures at the downstream end of the tunnel and the valve were derived for a system subjected to linear and slow valve closure. The analytical results were then compared with numerical ones obtained using the method of characteristics. There is agreement between them. The formulas can be applied to estimating water hammer pressure at the valve and transmission of water hammer pressure through the surge chamber at the junction for a hydraulic pipe system with a surge chamber.

Conditioning of spent mercury by amalgamation

International Nuclear Information System (INIS)

Yim, S. P.; Shon, J. S.; An, B. G.; Lee, H. J.; Lee, J. W.; Ji, C. G.; Kim, S. H.; Yoon, J. H.; Yang, M. S.

2002-01-01

Solidification by amalgamation was performed to immobilize and stabilize the liquid spent mercury. First, the appropriate metal and alloy which can convert liquid mercury into a solid form of amalgam were selected through initial tests. The amalgam form, formulated in optimum composition, was characterized and subjected to performance tests including compressive strength, water immersion, leachability and initial vaporization rate to evaluate mechanical integrity, durability and leaching properties. Finally, bench scale amalgamation trial was conducted with about 1 kg of spent mercury to verify the feasibility of amalgamation method

Thermal analysis of a one-element PWR spent fuel shipping cask

International Nuclear Information System (INIS)

Fields, S.R.

1979-06-01

The transient thermal behavior of a typical one-element PWR spent fuel shipping cask, following a hypothetical accident and fire, has been simulated. The objectives of the study were to determine the transient behavior of the cask and its spent fuel primary coolant through the pressure relief system and possible fuel pin clad failure due to overheating following loss of coolant. 15 figures, 7 tables

Evaluation of the heat transfer in a geological repository concept containing PWR, VHTR and hybrid ads-fission spent fuels

Energy Technology Data Exchange (ETDEWEB)

Jonusan, Raoni A.S.; Pereira, Fernando; Velasquez, Carlos E.; Salome, Jean A.D.; Cardoso, Fabiano; Pereira, Claubia; Fortini, Angela, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-11-01

The investigation of the thermal behavior of spent fuel (SF) materials is essential to determining appropriate potential sites to accommodate geological repositories as well as the design of canisters, considering their potential risk to people health and of environmental contamination. This work presents studies of the temperature in a canister containing spent fuels discharged from Pressurized Water Reactor (PWR), Very High-Temperature Reactor (VHTR) and Accelerator-Driven Subcritical Reactor System (ADS) reactor systems in a geological repository concept. The thermal analyses were performed with the software ANSYS, which is widely used to solve engineering problems through the Finite Element Method. The ANSYS Transient Thermal module was used. The spent nuclear fuels were set as heat sources using data of previous studies derived from decay heat curves. The studies were based on comparison of the mean temperature on a canister surface along the time under geological disposal conditions, for a same amount of each type of spent nuclear fuel evaluated. The results conclude that fuels from VHTR and ADS systems are inappropriate to be disposed in a standardized PWR canister, demanding new studies to determine the optimal amount of spent fuel and new internal canister geometries. It is also possible to conclude that the hypothetical situation of a single type of canister being used to accommodate different types of spent nuclear fuels is not technically feasible. (author)

High pressure water jet cutting and stripping

Science.gov (United States)

Hoppe, David T.; Babai, Majid K.

1991-01-01

High pressure water cutting techniques have a wide range of applications to the American space effort. Hydroblasting techniques are commonly used during the refurbishment of the reusable solid rocket motors. The process can be controlled to strip a thermal protective ablator without incurring any damage to the painted surface underneath by using a variation of possible parameters. Hydroblasting is a technique which is easily automated. Automation removes personnel from the hostile environment of the high pressure water. Computer controlled robots can perform the same task in a fraction of the time that would be required by manual operation.

Decontamination and recycle of zirconium pressure tubes from Pressurized Heavy Water Reactor

International Nuclear Information System (INIS)

Gantayet, L.M.; Verma, R.; Remya Devi, P.S.; Banerjee, S.; Kotak, V.; Raha, A.; Sandeep, K.C.; Joshi, Shreeram W.; Lali, A.M.

2009-01-01

An ion exchange process has been developed for decontamination of zirconium pressure tubes from Pressurized Heavy Water Reactor and recycling of neutronically improved zirconium. Distribution coefficient, equilibrium isotherm, kinetic and breakthrough data were used to develop the separation process. Effect of gamma radiation on indigenous resins was also studied to assess their suitability in high radiation field. (author)

Structure and dynamics of water confined in a graphene nanochannel under gigapascal high pressure: dependence of friction on pressure and confinement.

Science.gov (United States)

Yang, Lei; Guo, Yanjie; Diao, Dongfeng

2017-05-31

Recently, water flow confined in nanochannels has become an interesting topic due to its unique properties and potential applications in nanofluidic devices. The trapped water is predicted to experience high pressure in the gigapascal regime. Theoretical and experimental studies have reported various novel structures of the confined water under high pressure. However, the role of this high pressure on the dynamic properties of water has not been elucidated to date. In the present study, the structure evolution and interfacial friction behavior of water constrained in a graphene nanochannel were investigated via molecular dynamics simulations. Transitions of the confined water to different ice phases at room temperature were observed in the presence of lateral pressure at the gigapascal level. The friction coefficient at the water/graphene interface was found to be dependent on the lateral pressure and nanochannel height. Further theoretical analyses indicate that the pressure dependence of friction is related to the pressure-induced change in the structure of water and the confinement dependence results from the variation in the water/graphene interaction energy barrier. These findings provide a basic understanding of the dynamics of the nanoconfined water, which is crucial in both fundamental and applied science.

Short-term storage considerations for spent plutonium-thorium fuel bundles

Energy Technology Data Exchange (ETDEWEB)

Blomeley, L.; Dugal, C.; Masala, E.; Tran, T., E-mail: laura.blomeley@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2015-12-15

To support the development of advanced pressurized heavy water reactor (PHWR) fuel cycles, it is necessary to study short-term storage solutions for spent reactor fuel. In this paper, some representational criticality safety and shielding assessments are presented for a particular PHWR plutonium-thorium based fuel bundle concept in a hypothetical aboveground dry storage module. The criticality assessment found that the important parameters for the storage design are neutron absorber content and fuel composition, particularly in light of the high sensitivity of code results to plutonium. The shielding assessment showed that the shielding as presented in the paper would need to be redesigned to provide greater gamma attenuation. These findings can be used to aid in designing fuel storage facilities. (author)

International Experts’ Meeting on Reactor and Spent Fuel Safety in the Light of the Accident at the Fukushima Daiichi Nuclear Power Plant. Presentations

International Nuclear Information System (INIS)

2012-01-01

The primary objectives of this International Experts’ Meeting (IEM) were: to analyse relevant technical aspects of reactor and spent nuclear fuel management safety and performance related to severe accidents; to review what is known to date about the accident at the Fukushima Daiichi nuclear power plant in order to understand more fully its root causes; and to share the lessons learned from the accident. The meeting identified the necessary priorities for further actions in these areas in different power reactor types, focusing in particular on boiling water reactors (BWRs) and pressurized water reactors (PWRs). The meeting provided a forum for discussions and exchange of information among technical experts from Member States on reactor and spent nuclear fuel safety and performance under severe conditions. The meeting was of particular interest to technical experts from utilities, research and design organizations, regulatory bodies, manufacturing and service companies and other stakeholders. In particular, the objectives of the meeting was to: • Identify and analyse reactor and spent nuclear fuel safety and performance issues; • Consider the design, engineering and analysis of current and new systems for accident prevention and mitigation; • Exchange information on national assessments of reactor and spent nuclear fuel safety and performance; and • Identify potential priority areas for research and development, technology development and management

Initiation of depleted uranium oxide and spent fuel testing for the spent fuel sabotage aerosol ratio program

Energy Technology Data Exchange (ETDEWEB)

Molecke, M.A.; Gregson, M.W.; Sorenson, K.B. [Sandia National Labs. (United States); Billone, M.C.; Tsai, H. [Argonne National Lab. (United States); Koch, W.; Nolte, O. [Fraunhofer Inst. fuer Toxikologie und Experimentelle Medizin (Germany); Pretzsch, G.; Lange, F. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (Germany); Autrusson, B.; Loiseau, O. [Inst. de Radioprotection et de Surete Nucleaire (France); Thompson, N.S.; Hibbs, R.S. [U.S. Dept. of Energy (United States); Young, F.I.; Mo, T. [U.S. Nuclear Regulatory Commission (United States)

2004-07-01

We provide a detailed overview of an ongoing, multinational test program that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolized materials plus volatilized fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high energy density device, HEDD. The program participants in the U.S. plus Germany, France, and the U.K., part of the international Working Group for Sabotage Concerns of Transport and Storage Casks, WGSTSC have strongly supported and coordinated this research program. Sandia National Laboratories, SNL, has the lead role for conducting this research program; test program support is provided by both the U.S. Department of Energy and Nuclear Regulatory Commission. WGSTSC partners need this research to better understand potential radiological impacts from sabotage of nuclear material shipments and storage casks, and to support subsequent risk assessments, modeling, and preventative measures. We provide a summary of the overall, multi-phase test design and a description of all explosive containment and aerosol collection test components used. We focus on the recently initiated tests on ''surrogate'' spent fuel, unirradiated depleted uranium oxide, and forthcoming actual spent fuel tests. The depleted uranium oxide test rodlets were prepared by the Institut de Radioprotection et de Surete Nucleaire, in France. These surrogate test rodlets closely match the diameter of the test rodlets of actual spent fuel from the H.B. Robinson reactor (high burnup PWR fuel) and the Surry reactor (lower, medium burnup PWR fuel), generated from U.S. reactors. The characterization of the spent fuels and fabrication into short, pressurized rodlets has been performed by Argonne National Laboratory, for testing at SNL. The ratio of the aerosol and respirable particles released from HEDD-impacted spent

Cavitation nuclei in water exposed to transient pressures

DEFF Research Database (Denmark)

Andersen, Anders Peter; Mørch, Knud Aage

2015-01-01

A model of skin-stabilized interfacial cavitation nuclei and their response to tensile and compressive stressing is presented. The model is evaluated in relation to experimental tensile strength results for water at rest at the bottom of an open water-filled container at atmospheric pressure...... and room temperature. These results are obtained by recording the initial growth of cavities generated by a short tensile pulse applied to the bottom of the container. It is found that the cavitation nuclei shift their tensile strength depending on their pressure history. Static pressurization...... for an extended period of time prior to testing is known to increase the tensile strength of water, but little information is available on how it is affected by compression pulses of short duration. This is addressed by imposing compression pulses of approximately 1 ms duration and a peak intensity of a few bar...

Preliminary assessment of alternative dry storage methods for the storage of commercial spent nuclear fuel

International Nuclear Information System (INIS)

1981-11-01

This report presents the results of an assessment of the (1) state of technology, (2) licensability, (3) implementation schedule, and (4) costs of alternative dry methods for storage of spent fuel at a reactor location when used to supplement reactor pool storage facilities. The methods of storage that were considered included storage in casks, drywells, concrete silos and air-cooled vaults. The impact of disassembly of spent fuel and storage of consolidated fuel rods was also determined. The economic assessments were made based on the current projected storage requirements of Virginia Electric and Power Company's Surry Station for the period 1985 to 2009, which has two operating pressurized water reactors (824 MWe each). It was estimated that the unit cost for storage of spent fuel in casks would amount to $117/kgU and that such costs for storage in drywells would amount to $137/kgU. However, based on the overall assessment it was concluded both storage methods were equal in merit. Modular methods of storage were generally found to be more economic than those requiring all or most of the facilities to be constructed prior to commencement of storage operations

Water cycles in closed ecological systems: effects of atmospheric pressure.

Science.gov (United States)

Rygalov, Vadim Y; Fowler, Philip A; Metz, Joannah M; Wheeler, Raymond M; Bucklin, Ray A

2002-01-01

In bioregenerative life support systems that use plants to generate food and oxygen, the largest mass flux between the plants and their surrounding environment will be water. This water cycle is a consequence of the continuous change of state (evaporation-condensation) from liquid to gas through the process of transpiration and the need to transfer heat (cool) and dehumidify the plant growth chamber. Evapotranspiration rates for full plant canopies can range from ~1 to 10 L m-2 d-1 (~1 to 10 mm m-2 d-1), with the rates depending primarily on the vapor pressure deficit (VPD) between the leaves and the air inside the plant growth chamber. VPD in turn is dependent on the air temperature, leaf temperature, and current value of relative humidity (RH). Concepts for developing closed plant growth systems, such as greenhouses for Mars, have been discussed for many years and the feasibility of such systems will depend on the overall system costs and reliability. One approach for reducing system costs would be to reduce the operating pressure within the greenhouse to reduce structural mass and gas leakage. But managing plant growth environments at low pressures (e.g., controlling humidity and heat exchange) may be difficult, and the effects of low-pressure environments on plant growth and system water cycling need further study. We present experimental evidence to show that water saturation pressures in air under isothermal conditions are only slightly affected by total pressure, but the overall water flux from evaporating surfaces can increase as pressure decreases. Mathematical models describing these observations are presented, along with discussion of the importance for considering "water cycles" in closed bioregenerative life support systems.

Water cycles in closed ecological systems: effects of atmospheric pressure

Science.gov (United States)

Rygalov, Vadim Y.; Fowler, Philip A.; Metz, Joannah M.; Wheeler, Raymond M.; Bucklin, Ray A.; Sager, J. C. (Principal Investigator)

2002-01-01

In bioregenerative life support systems that use plants to generate food and oxygen, the largest mass flux between the plants and their surrounding environment will be water. This water cycle is a consequence of the continuous change of state (evaporation-condensation) from liquid to gas through the process of transpiration and the need to transfer heat (cool) and dehumidify the plant growth chamber. Evapotranspiration rates for full plant canopies can range from 1 to 10 L m-2 d-1 (1 to 10 mm m-2 d-1), with the rates depending primarily on the vapor pressure deficit (VPD) between the leaves and the air inside the plant growth chamber. VPD in turn is dependent on the air temperature, leaf temperature, and current value of relative humidity (RH). Concepts for developing closed plant growth systems, such as greenhouses for Mars, have been discussed for many years and the feasibility of such systems will depend on the overall system costs and reliability. One approach for reducing system costs would be to reduce the operating pressure within the greenhouse to reduce structural mass and gas leakage. But managing plant growth environments at low pressures (e.g., controlling humidity and heat exchange) may be difficult, and the effects of low-pressure environments on plant growth and system water cycling need further study. We present experimental evidence to show that water saturation pressures in air under isothermal conditions are only slightly affected by total pressure, but the overall water flux from evaporating surfaces can increase as pressure decreases. Mathematical models describing these observations are presented, along with discussion of the importance for considering "water cycles" in closed bioregenerative life support systems.

Physical and decay characteristics of commercial LWR spent fuel

International Nuclear Information System (INIS)

Roddy, J.W.; Claiborne, H.C.; Ashline, R.C.; Johnson, P.J.; Rhyne, B.T.

1986-01-01

Information was collected from the literature and from major manufacturers that will be useful in the design and construction of a mined geologic repository for the disposal of light-water-reactor spent fuel. Pertinent data are included on mechanical design characteristics and materials of construction for fuel assemblies and fuel rods and computed values for heat generation rates, radioactivity, and photon and neutron emission rates as a function of time for four reference cases. Calculations were made with the ORIGEN2 computer code for burnups of 27,500 and 40,000 MWd for a typical boiling-water reactor and 33,000 and 60,000 MWd for a typical pressurized-water reactor. The results are presented in figures depicting the individual contributions per metric ton of initial heavy metal for the activation products, fission products, and actinides and their daughters to the radioactivity and thermal power as a function of time. Tables are also presented that list the contribution of each major nuclide to the radioactivity, thermal power, and photons and neutrons emitted for disposal emitted for disposal periods from 1 to 100,000 years

Seismic evaluation of BWR spent fuel storage racks using actual damping by vibration test in water

International Nuclear Information System (INIS)

Yamasaki, Hiroto; Iwakura, Shigeyoshi; Imaoka, Tetsuo; Okumura, Kazue; Orita, Syuichi; Namita, Yoshio

2010-01-01

Damping value for BWR spent fuel storage racks has been used 1 percent damping, which is applied to welded steel structures in air as defined JEAG4601. However, it is considered that the actual damping is higher than that of the above mentioned, because of its underwater installation. This report shows the actual damping value of the Check Arrayed Rack by vibration test in water and Evaluation by the analysis of rack using actual damping. (author)

Development of code SFINEL (Spent fuel integrity evaluator)

Energy Technology Data Exchange (ETDEWEB)

Kim, Yong Soo; Min, Chin Young; Ohk, Young Kil; Yang, Yong Sik; Kim, Dong Ju; Kim, Nam Ku [Hanyang University, Seoul (Korea)

1999-01-01

SFINEL code, an integrated computer program for predicting the spent fuel rod integrity based on burn-up history and major degradation mechanisms, has been developed through this project. This code can sufficiently simulate the power history of a fuel rod during the reactor operation and estimate the degree of deterioration of spent fuel cladding using the recently-developed models on the degradation mechanisms. SFINEL code has been thoroughly benchmarked against the collected in-pile data and operating experiences: deformation and rupture, and cladding oxidation, rod internal pressure creep, then comprehensive whole degradation process. (author). 75 refs., 51 figs., 5 tabs.

Emergency water supply facility for nuclear reactor

International Nuclear Information System (INIS)

Karasawa, Toru

1998-01-01

Water is stored previously in an equipment storage pit disposed on an operator floor of a reactor building instead of a condensate storage vessel. Upon occurrence of an emergency, water is supplied from the equipment storage pit by way of a sucking pipeline to a pump of a high pressure reactor core water injection circuit and a pump of a reactor-isolation cooling circuit to supply water to a reactor. The equipment storage pit is arranged in a building so that the depth thereof is determined to keep the required amount of water by storing water at a level lower than the lower end of a pool gate during normal operation. Water is also supplied from the equipment storage pit by way of a supply pipeline to a spent fuel storage pool on the operation floor of the reactor building. Namely, water is supplied to the spent fuel storage pool by a pump of a fuel pool cooling and cleaning circuit. This can eliminate a suppression pool cleaning circuit. (I.N.)

Data book of the isotopic composition of spent fuel in light water reactors

International Nuclear Information System (INIS)

Naito, Yoshitaka; Kurosawa, Masayoshi; Kaneko, Toshiyuki.

1994-03-01

In the framework of the activity of the working group on Evaluation of Nuclide Generation and Depletion in the Japanese Nuclear Data Committee, we summarized the assay data of the isotopic composition of LWR spent fuels in order to verify the accuracy of the burnup calculation codes. The report contains the data collected from the 13 light water reactors (LWRs) including the 9 LWRs (5 PWRs and 4 BWRs) in Europe and USA, the 4 LWRs (2 PWRs and 2 BWRs) in Japan. The collected data were sorted into the irradiation history of the fuel samples, the composition of the fuel assemblies, the sampling position and the isotopic composition of the fuel samples. (author)

Variation of Pore Water Pressure in Tailing Sand under Dynamic Loading

Directory of Open Access Journals (Sweden)

Jia-xu Jin

2018-01-01

Full Text Available Intense vibration affects the pore water pressure in a tailing dam, with the tendency to induce dam liquefaction. In this study, experiments were performed wherein model tailing dams were completely liquefied by sustained horizontal dynamic loading to determine the effects of the vibration frequency, vibration amplitude, and tailing density on the pore water pressure. The results revealed four stages in the increase of the tailing pore water pressure under dynamic loading, namely, a slow increase, a rapid increase, inducement of structural failure, and inducement of complete liquefaction. A lower frequency and smaller amplitude of the vibration were found to increase the time required to achieve a given pore water pressure in dense tailings. Under the effect of these three factors—vibration frequency and amplitude and tailing density—the tailing liquefaction time varied nonlinearly with the height from the base of the tailing dam, with an initial decrease followed by an increase. The pore pressure that induced structural failure also gradually decreased with increasing height. The increase in the tailing pore pressure could be described by an S-shaped model. A complementary multivariate nonlinear equation was also derived for predicting the tailing pore water pressure under dynamic loading.

The pressurized water reactor

International Nuclear Information System (INIS)

Gallagher, J.L.

1987-01-01

Pressurized water reactor technology has reached a maturity that has engendered a new surge of innovation, which in turn, has led to significant advances in the technology. These advances, characterized by bold thinking but conservative execution, are resulting in nuclear plant designs which offer significant performance and safety improvements. This paper describes the innovations which are being designed into mainstream PWR technology as well as the desings which are resulting from such innovations. (author)

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)

Study on low pressure evaporation of fresh water generation system model

Energy Technology Data Exchange (ETDEWEB)

Chung, Han Shik; Wibowo, Supriyanto; Shin, Yong Han; Jeong, Hyo Min [Gyeongsang National University, Tongyeong (Korea, Republic of); Fajar, Berkah [University of Diponegoro, Semarang (Indonesia)

2012-02-15

A low pressure evaporation fresh water generation system is designed for converting brackish water or seawater into fresh water by distillation in low pressure and temperature. Distillation through evaporation of feed water and subsequent vapor condensation as evaporation produced fresh water were studied; tap water was employed as feed water. The system uses the ejector as a vacuum creator of the evaporator, which is one of the most important parts in the distillation process. Hence liquid can be evaporated at a lower temperature than at normal or atmospheric conditions. Various operating conditions, i.e. temperature of feed water and different orifice diameters, were applied in the experiment to investigate the characteristics of the system. It was found that these parameters have a significant effect on the performance of fresh water generation systems with low pressure evaporation.

Separation and Recovery of Uranium Metal from Spent Light Water Reactor Fuel via Electrolytic Reduction and Electrorefining

International Nuclear Information System (INIS)

Herrmann, S.D.; Li, S.X.

2010-01-01

A series of bench-scale experiments was performed in a hot cell at Idaho National Laboratory to demonstrate the separation and recovery of uranium metal from spent light water reactor (LWR) oxide fuel. The experiments involved crushing spent LWR fuel to particulate and separating it from its cladding. Oxide fuel particulate was then converted to metal in a series of six electrolytic reduction runs that were performed in succession with a single salt loading of molten LiCl - 1 wt% Li2O at 650 C. Analysis of salt samples following the series of electrolytic reduction runs identified the diffusion of select fission products from the spent fuel to the molten salt electrolyte. The extents of metal oxide conversion in the post-test fuel were also quantified, including a nominal 99.7% conversion of uranium oxide to metal. Uranium metal was then separated from the reduced LWR fuel in a series of six electrorefining runs that were performed in succession with a single salt loading of molten LiCl-KCl-UCl3 at 500 C. Analysis of salt samples following the series of electrorefining runs identified additional partitioning of fission products into the molten salt electrolyte. Analyses of the separated uranium metal were performed, and its decontamination factors were determined.

Nonlinear vibration of a hemispherical dome under external water pressure

International Nuclear Information System (INIS)

Ross, C T F; McLennan, A; Little, A P F

2011-01-01

The aim of this study was to analyse the behaviour of a hemi-spherical dome when vibrated under external water pressure, using the commercial computer package ANSYS 11.0. In order to achieve this aim, the dome was modelled and vibrated in air and then in water, before finally being vibrated under external water pressure. The results collected during each of the analyses were compared to the previous studies, and this demonstrated that ANSYS was a suitable program and produced accurate results for this type of analysis, together with excellent graphical displays. The analysis under external water pressure, clearly demonstrated that as external water pressure was increased, the resonant frequencies decreased and a type of dynamic buckling became likely; because the static buckling eigenmode was similar to the vibration eigenmode. ANSYS compared favourably with the in-house software, but had the advantage that it produced graphical displays. This also led to the identification of previously undetected meridional modes of vibration; which were not detected with the in-house software.

Nonlinear vibration of a hemispherical dome under external water pressure

Science.gov (United States)

Ross, C. T. F.; McLennan, A.; Little, A. P. F.

2011-07-01

The aim of this study was to analyse the behaviour of a hemi-spherical dome when vibrated under external water pressure, using the commercial computer package ANSYS 11.0. In order to achieve this aim, the dome was modelled and vibrated in air and then in water, before finally being vibrated under external water pressure. The results collected during each of the analyses were compared to the previous studies, and this demonstrated that ANSYS was a suitable program and produced accurate results for this type of analysis, together with excellent graphical displays. The analysis under external water pressure, clearly demonstrated that as external water pressure was increased, the resonant frequencies decreased and a type of dynamic buckling became likely; because the static buckling eigenmode was similar to the vibration eigenmode. ANSYS compared favourably with the in-house software, but had the advantage that it produced graphical displays. This also led to the identification of previously undetected meridional modes of vibration; which were not detected with the in-house software.

Studies on recycling and utilization of spent catalysts. Preparation of active hydrodemetallization catalyst compositions from spent residue hydroprocessing catalysts

Energy Technology Data Exchange (ETDEWEB)

Marafi, Meena; Stanislaus, Antony [Petroleum Refining Department, Petroleum Research and Studies Center, Kuwait Institute for Scientific Research, P.O. Box 24885, Safat (Kuwait)

2007-02-15

Spent catalysts form a major source of solid wastes in the petroleum refining industries. Due to environmental concerns, increasing emphasis has been placed on the development of recycling processes for the waste catalyst materials as much as possible. In the present study the potential reuse of spent catalysts in the preparation of active new catalysts for residual oil hydrotreating was examined. A series of catalysts were prepared by mixing and extruding spent residue hydroprocessing catalysts that contained C, V, Mo, Ni and Al{sub 2}O{sub 3} with boehmite in different proportions. All prepared catalysts were characterized by chemical analysis and by surface area, pore volume, pore size and crushing strength measurements. The hydrodesulfurization (HDS) and hydrodemetallization (HDM) activities of the catalysts were evaluated by testing in a high pressure fixed-bed microreactor unit using Kuwait atmospheric residue as feed. A commercial HDM catalyst was also tested under similar operating conditions and their HDS and HDM activities were compared with that of the prepared catalysts. The results revealed that catalyst prepared with addition of up to 40 wt% spent catalyst to boehmite had fairly high surface area and pore volume together with large pores. The catalyst prepared by mixing and extruding about 40 wt% spent catalyst with boehmite was relatively more active for promoting HDM and HDS reactions than a reference commercial HDM catalyst. The formation of some kind of new active sites from the metals (V, Mo and Ni) present in the spent catalyst is suggested to be responsible for the high HDM activity of the prepared catalyst. (author)

Water chemistry management of research reactor in JAERI

Energy Technology Data Exchange (ETDEWEB)

Yoshijima, Tetsuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-10-01

The JRR-3M cooling system consists of four systems, namely; (1) primary cooling system, (2) heavy water cooling system, (3) helium system and (4) secondary cooling system. The heavy water is used for reflector and pressurized with helium gas. Water chemistry management of the JRR-3M cooling systems is one of the important subject for the safety operation. The main objects are to prevent the corrosion of cooling system and fuel elements, to suppress the plant radiation build-up and to minimize the generation of radioactive waste. All measured values were within the limits of specifications and JRR-3M reactor was operated with safety in 1996. Spent fuels of JRR-3M reactor are stored in the spent fuel pool. This pool water has been analyzed to prevent corrosion of aluminum cladding of spent fuels. Water chemistry of spent fuel pool water is applied to the prevention of corrosion of aluminum alloys including fuel cladding. The JRR-2 reactor was eternally stopped in December 1996 and is now under decommissioning. The JRR-2 reactor is composed of heavy water tank, fuel guide tube and horizontal experimental hole. These are constructed of aluminum alloy and biological shield and upper shield are constructed of concrete. Three types of corrosion of aluminum alloy were observed in the JRR-2. The Alkaline corrosion of aluminum tube occurred in 1972 because of the mechanical damage of the aluminum fuel guide tube which is used for fuel handling. Modification of the reactor top shield was started in 1974 and completed in 1975. (author)

When immiscible becomes miscible-Methane in water at high pressures.

Science.gov (United States)

Pruteanu, Ciprian G; Ackland, Graeme J; Poon, Wilson C K; Loveday, John S

2017-08-01

At low pressures, the solubility of gases in liquids is governed by Henry's law, which states that the saturated solubility of a gas in a liquid is proportional to the partial pressure of the gas. As the pressure increases, most gases depart from this ideal behavior in a sublinear fashion, leveling off at pressures in the 1- to 5-kbar (0.1 to 0.5 GPa) range with solubilities of less than 1 mole percent (mol %). This contrasts strikingly with the well-known marked increase in solubility of simple gases in water at high temperature associated with the critical point (647 K and 212 bar). The solubility of the smallest hydrocarbon, the simple gas methane, in water under a range of pressure and temperature is of widespread importance, because it is a paradigmatic hydrophobe and occurs widely in terrestrial and extraterrestrial geology. We report measurements up to 3.5 GPa of the pressure dependence of the solubility of methane in water at 100°C-well below the latter's critical temperature. Our results reveal a marked increase in solubility between 1 and 2 GPa, leading to a state above 2 GPa where the maximum solubility of methane in water exceeds 35 mol %.

Standard Technical Specifications for Westinghouse pressurized water reactors

International Nuclear Information System (INIS)

Virgilio, M.J.

1980-09-01

The Standard Technical Specifications for Westinghouse Pressurized Water Reactors (W-STS) is a generic document prepared by the U.S. NRC for use in the licensing process of current Westinghouse Pressurized Water Reactors. The W-STS sets forth the Limits, Operating Conditions and other requirements applicable to nuclear reactor facility operation as set forth in by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. This document is revised periodically to reflect current licensing requirements

Flooding of a large, passive, pressure-tube light water reactor

International Nuclear Information System (INIS)

Hejzlar, P.; Todreas, N.E.; Driscoll, M.J.

1997-01-01

A reactor concept has been developed which can survive loss of coolant accidents without scram and without replenishing primary coolant inventory, while maintaining safe temperature limits on the fuel and pressure tubes. The proposed concept is a pressure tube type reactor of similar design to CANDU reactors, but differing in three key aspects. First, a solid SiC-coated graphite fuel matrix is used in place of fuel pin bundles to enable the dissipation of decay heat from the fuel in the absence of primary coolant. Second, the heavy water coolant in the pressure tubes is replaced by light water, which also serves as the moderator. Finally, the calandria tank, surrounded by a graphite reflector, contains a low pressure gas instead of heavy water moderator, and this normally-voided calandria is connected to a light water heat sink. The cover gas displaces the light water from the calandria during normal operation, while during loss of coolant or loss of heat sink accidents it allows passive calandria flooding. Calandria flooding also provides redundant and diverse reactor shutdown. This paper describes the thermal hydraulic characteristics of the passively initiated, gravity driven calandria flooding process. Flooding the calandria space with light water is a unique and very important feature of the proposed pressure-tube light water reactor (PTLWR) concept. The flooding of the top row of fuel channels must be accomplished fast enough so that in the total loss of coolant, none of the critical components of the fuel channel, i.e. the pressure tube, the calandria tube, the matrix and the fuel, exceed their design limits. The flooding process has been modeled and shown to be rapid enough to maintain all components within their design limits. (orig.)

Spent fuel storage rack

International Nuclear Information System (INIS)

Kurokawa, Hideaki; Kumagaya, Naomi; Oda, Masashi; Matsuda, Masami; Maruyama, Hiromi; Yamanaka, Tsuneyasu.

1997-01-01

The structure of a spent fuel storage rack is determined by the material, thickness, size of square cylindrical tubes (the gap between spent fuel assemblies and the square cylindrical tubes) and pitch of the arrangement (the gap between each of the square cylindrical tubes). In the present invention, the thickness and the pitch of the arrangement of the square tubes are optimized while evaluating subcriticality. Namely, when the sum of the thickness of the water gap at the outer side (the pitch of arrangement of the cylindrical tubes) and the thickness of the cylindrical tubes is made constant, the storage rack is formed by determining the thickness of the cylindrical tubes which is smaller than the optimum value among the combination of the thickness of the water gap at the outer side and that of the cylindrical tube under the effective multiplication factor to be performed. Then, the weight of the rack can be reduced, and the burden of the load on the bottom of the pool can be reduced. Further, the amount of the constitutional materials of the rack itself can be reduced thereby capable of reducing the cost for the materials of the rack. (T.M.)

Pumps for German pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Dernedde, R.

1984-01-01

The article describes the development of a selection of pumps which are used in the primary coolant system and the high-pressure safety injection system and feed water system during the past 2 decades. The modifications were caused by the step-wise increasing power output of the plants from 300 MW up to 1300 MW. Additional important influences were given be the increased requirements for quality assurance and final-documentation. The good operating results of the delivered pumps proved that the reliability is independent of the volume of the software-package. The outlook expects that consolidation will be followed by additional steps for the order processing of components for the convoy pumps. KW: main coolant pump; primary system; boiler feed pump; reactor pump; secondary system; barrel insert pump; pressure water reactor; convoy pump; state of the art.

Thermal Cooling Limits of Sbotaged Spent Fuel Pools

Energy Technology Data Exchange (ETDEWEB)

Dr. Thomas G. Hughes; Dr. Thomas F. Lin

2010-09-10

To develop the understanding and predictive measures of the post “loss of water inventory” hazardous conditions as a result of the natural and/or terrorist acts to the spent fuel pool of a nuclear plant. This includes the thermal cooling limits to the spent fuel assembly (before the onset of the zircaloy ignition and combustion), and the ignition, combustion, and the subsequent propagation of zircaloy fire from one fuel assembly to others

Materials behavior in interim storage of spent fuel

International Nuclear Information System (INIS)

Johnson, A.B. Jr.; Bailey, W.J.; Gilbert, E.R.; Inman, S.C.

1982-01-01

Interim storage has emerged as the only current spent-fuel management method in the US and is essential in all countries with nuclear reactors. Materials behavior is a key aspect in licensing interim-storage facilities for several decades of spent-fuel storage. This paper reviews materials behavior in wet storage, which is licensed for light-water reactor (LWR) fuel, and dry storage, for which a licensing position for LWR fuel is developing

Study on leakage rates of high temperature water from wet-type transport casks for spent fuel. Pt. 2. Leakage rates from a scratch on O-ring surface and narrow wires adhering to O-ring surface

International Nuclear Information System (INIS)

Asano, R.; Aritomi, M.; Sudi, A.; Kohketsu, Y.

1997-01-01

A programme for enhancement of fuel burnup has been promoted in Japan as part of the sophisticated programme for light water reactors to reduce the fuel cost and the amount of spent fuel. As part of this fuel programme, a new wet-type transport cask has been developed to transport the high burnup fuels efficiently. The purpose of this work is to clarify the margin of safety in the evaluation of the release rate of radioactive materials from the wet-type transport cask into the environment and to establish a practical evaluation method for leakage rates on leak behaviour of high temperature water from the casks. In this paper, leakage rates of water under high pressures and at high temperatures are investigated from two kinds of leak path model. One is a disc with a scratch on the surface which simulates a defect on the seal surface of the O-ring flange and the other is narrow stainless steel wires installed on the O-ring surface which simulates hair adhering to the O-ring surface. From the results, an evaluation method for the leakage rate of water under high pressure and at high temperature from a non-circular leak path and multiple leak paths is proposed. (author)

Research on axial total pressure distributions of sonic steam jet in subcooled water

International Nuclear Information System (INIS)

Wu Xinzhuang; Li Wenjun; Yan Junjie

2012-01-01

The axial total pressure distributions of sonic steam jet in subcooled water were experimentally investigated for three different nozzle diameters (6.0 mm, 8.0 mm and 10.0 mm). The inlet steam pressure, and pool subcooling subcooled water temperature were in the range of 0.2-0.6 MPa and 420-860 ℃, respectively. The effect of steam pressure, subcooling water temperature and nozzle size on the axial pressure distributions were obtained, and also the characteristics of the maximum pressure and its position were studied. The results indicated that the characteristics of the maximum pressure were influenced by the nozzle size for low steam pressure, but the influence could be ignored for high steam pressure. Moreover, a correlation was given to correlate the position of the maximum pressure based on steam pressure and subcooling water temperature, and the discrepancies of predictions and experiments are within ±15%. (authors)

The effect of pressurizer-water-level on the low frequency component of the pressure spectrum in a PWR

International Nuclear Information System (INIS)

Por, G.; Izsak, E.; Valko, J.

1984-09-01

The pressure fluctuations were measured in the cooling system of the Paks-1 reactor. A shift of the peak was detected in low frequency component of the pressure fluctuation spectrum which is due to the fluctuations of water level in the pressurizer. Using the model of Katona and Nagy (1983), the eigenfrequencies, the magnitude of the shift and the sensitivity to the pressurizer water level were reproduced in good accordance with the experimental data. (D.Gy.)

Particle Size Effects on Fenton Regeneration of MTBE-spent Activated Carbon

Science.gov (United States)

Fenton-driven regeneration of spent granular activated carbon (GAC) is a developing technology that may reduce water treatment costs. In this study, the effect of GAC particle size on Fenton-driven oxidation of methyl tert-butyl ether (MTBE)-spent GAC was evaluated. The GAC was...

BNFL's new spent fuel transport flask - Excellox 8

International Nuclear Information System (INIS)

McWilliam, D.S.

2002-01-01

Since British Nuclear Fuels plc (BNFL) was formed in 1971 its transport service has safely moved spent light water reactor fuel from many locations abroad to its fuel handling plants at Sellafield in the UK. To support this business a number of types of flasks have been designed and used. One of the types used has been the Excellox family of water-filled flasks. To support future business opportunities a new flask, designed to meet the requirements of the new IAEA transport regulations TS-R-1 (ST-1, Revised), has been developed. The flask will be a type B(U)F. This new flask design will maximise fuel carrying capacity to minimise transport costs. The design capacity of the new Excellox 8 flask is to be 12 pressurised water reactor or 32 boiling water reactor fuel assemblies. The objective of this BNFL project is to provide another economic spent nuclear fuel transport system, in support of BNFL transport business. (author)

Type GQS-1 high pressure steam manifold water level monitoring system

International Nuclear Information System (INIS)

Li Nianzu; Li Beicheng; Jia Shengming

1993-10-01

The GQS-1 high pressure steam manifold water level monitoring system is an advanced nuclear gauge that is suitable for on-line detecting and monitor in high pressure steam manifold water level. The physical variable of water level is transformed into electrical pulses by the nuclear sensor. A computer is equipped for data acquisition, analysis and processing and the results are displayed on a 14 inch color monitor. In addition, a 4 ∼ 20 mA output current is used for the recording and regulation of water level. The main application of this gauge is for on-line measurement of high pressure steam manifold water level in fossil-fired power plant and other industries

Evaluation of pressure transducers under turbid natural waters

Digital Repository Service at National Institute of Oceanography (India)

Joseph, A.; Desa, E.; Desa, E.; Smith, D.; Peshwe, V.B.; VijayKumar, K.; Desa, J.A.E.

, the use of rho sub(eff) in contrast to the bulk density, significantly improves the measurement accuracy. For celar waters, precision density measurements made on discrete water samples agreed with rho sub(eff) values derived from pressure measurements...

Water pressure and ground vibrations induced by water guns near Bandon Road Lock and Dam and Lemont, Illinois

Science.gov (United States)

Adams, Ryan F.; Koebel, Carolyn M.; Morrow, William S.

2018-02-13

Multiple geophysical sensors were used to characterize the underwater pressure field and ground vibrations of a seismic water gun and its suitability to deter the movement of Asian carps (particularly the silver [Hypophthalmichthys molitrix] and bighead [Hypophthalmichthys nobilis] carps) while ensuring the integrity of surrounding structures. The sensors used to collect this information were blast-rated hydrophones, surface- and borehole-mounted geophones, and fixed accelerometers.Results from two separate studies are discussed in this report. The Brandon Road study took place in May 2014, in the Des Plaines River, in a concrete-walled channel downstream of the Brandon Road Lock and Dam near Joliet, Illinois. The Lemont study took place in June 2014, in a segment of the dolomite setblock-lined Chicago Sanitary and Ship Canal near Lemont, Illinois.Two criteria were evaluated to assess the potential deterrence to carp migration, and to minimize the expected effect on nearby structures from discharge of the seismic water gun. The first criterion was a 5-pound-per-square-inch (lb/in2) limit for dynamic underwater pressure variations. The second criterion was a maximum velocity and acceleration disturbance of 0.75 inch per second (in/s) for sensitive machinery (such as the lock gates and pumps) and 2.0 in/s adjacent to canal walls, respectively. The criteria were based on previous studies of fish responses to dynamic pressure variations, and effects of vibrations on the structural integrity of concrete walls.The Brandon Road study evaluated the magnitude and extent of the pressure field created by two water gun configurations in the concrete-walled channel downstream of the lock where channel depths ranged from 11 to 14 feet (ft). Data from a single 80-cubic-inch (in³) water gun set at 6 ft below water surface (bws) produced a roughly cylindrical 5-lb/in2 pressure field 20 ft in radius, oriented vertically, with the radius decreasing to less than 15 ft at the water

SCALE-4 analysis of pressurized water reactor critical configurations. Volume 1: Summary

International Nuclear Information System (INIS)

DeHart, M.D.

1995-03-01

The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original fresh composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized water reactors (PWR). The analysis methodology utilized for all calculations in this report is based on the modules and data associated with the SCALE-4 code system. Each of the five volumes comprising this report provides an overview of the methodology applied. Subsequent volumes also describe in detail the approach taken in performing criticality calculations for these PWR configurations: Volume 2 describes criticality calculations for the Tennessee Valley Authority's Sequoyah Unit 2 reactor for Cycle 3; Volume 3 documents the analysis of Virginia Power's Surry Unit 1 reactor for the Cycle 2 core; Volume 4 documents the calculations performed based on GPU Nuclear Corporation's Three Mile Island Unit 1 Cycle 5 core; and, lastly, Volume 5 describes the analysis of Virginia Power's North Anna Unit 1 Cycle 5 core. Each of the reactor-specific volumes provides the details of calculations performed to determine the effective multiplication factor for each reactor core for one or more critical configurations using the SCALE-4 system; these results are summarized in this volume. Differences between the core designs and their possible impact on the criticality calculations are also discussed. Finally, results are presented for additional analyses performed to verify that solutions were sufficiently converged

The design of the DUPIC spent fuel bundle counter

International Nuclear Information System (INIS)

Menlove, H.O.; Rinard, P.M.; Kroncke, K.E.; Lee, Y.G.

1997-05-01

A neutron coincidence detector had been designed to measure the amount of curium in the fuel bundles and associated process samples used in the direct use of plutonium in Canadian deuterium-uranium (CANDU) fuel cycle. All of the sample categories are highly radioactive from the fission products contained in the pressurized water reactor (PWR) spent fuel feed stock. Substantial shielding is required to protect the He-3 detectors from the intense gamma rays. The Monte Carlo neutron and photon calculational code has been used to design the counter with a uniform response profile along the length of the CANDU-type fuel bundle. Other samples, including cut PWR rods, process powder, waste, and finished rods, can be measured in the system. This report describes the performance characteristics of the counter and support electronics. 3 refs., 23 figs., 6 tabs

Chloride Ingress into Concrete under Water Pressure

DEFF Research Database (Denmark)

Lund, Mia Schou; Sander, Lotte Braad; Grelk, Bent

2011-01-01

The chloride ingress into concrete under water pressures of 100 kPa and 800 kPa have been investigated by experiments. The specimens were exposed to a 10% NaCl solution and water mixture. For the concrete having w/c = 0.35 the experimental results show the chloride diffusion coefficient at 800 k......Pa (～8 atm.) is 12 times greater than at 100 kPa (～1 atm.). For w/c = 0.45 and w/c = 0.55 the chloride diffusion coefficients are 7 and 3 times greater. This means that a change in pressure highly influences the chloride ingress into the concrete and thereby the life length models for concrete structures....

Storage of spent fuel from power reactors in India management and experience

International Nuclear Information System (INIS)

Changrani, R.D.; Bajpai, D.D.; Kodilkar, S.S.

1999-01-01

The spent fuel management programme in India is based on closing the nuclear fuel cycle with reprocessing option. This will enable the country to enhance energy security through maximizing utilization of available limited uranium resources while pursuing its Three Stage Nuclear Power Programme. Storage of spent fuel in water pools remains as prevailing mode in the near term. In view of inventory build up of spent fuel, an Away-From-Reactor (AFR) On-Site (OS) spent fuel storage facility has been made operational at Tarapur. Dry storage casks also have been developed as 'add on' system for additional storage of spent fuels. The paper describes the status and experience pertaining to spent fuel storage practices in India. (author)

Clean-up system for pool water in pressure suppression chamber and operation method therefor

Energy Technology Data Exchange (ETDEWEB)

Hirabayashi, Kentaro; Kinoshita, Shoichiro

1996-09-17

Pool water in a pressure suppression chamber of a BWR type reactor is sucked by a pump of an after-heat removing system. The pool water pressurized here is sent to the pressure suppression chamber by way of a heat exchanger and a test line backwarding pipeline to stir the pool water in the pressure suppression chamber. Further, the pool water pressurized by the pump is sent to the pressure suppression chamber by way of a filtration desalting device and an exit pipe to purify the pool water. Upon cleaning of pipelines before the start of a periodical test, the pool water sucked by the pump is sent to the filtration desalting device and recovered to the pressure suppression chamber. This can reduce the amount of impurities carried to the suppression chamber. After the cleaning of the pipelines, pool water is passed through the test line backwarding pipeline, so that the pool water can be stirred at the same time. (I.N.)

Pressure probe study of the water relations of Phycomyces blakesleeanus sporangiophores

Science.gov (United States)

Cosgrove, D. J.; Ortega, J. K.; Shropshire, W. Jr

1987-01-01

The physical characteristics which govern the water relations of the giant-celled sporangiophore of Phycomyces blakesleeanus were measured with the pressure probe technique and with nanoliter osmometry. These properties are important because they govern water uptake associated with cell growth and because they may influence expansion of the sporangiophore wall. Turgor pressure ranged from 1.1 to 6.6 bars (mean = 4.1 bars), and was the same for stage I and stage IV sporangiophores. Sporangiophore osmotic pressure averaged 11.5 bars. From the difference between cell osmotic pressure and turgor pressure, the average water potential of the sporangiophore was calculated to be about -7.4 bars. When sporangiophores were submerged under water, turgor remained nearly constant. We propose that the low cell turgor pressure is due to solutes in the cell wall solution, i.e., between the cuticle and the plasma membrane. Membrane hydraulic conductivity averaged 4.6 x 10(-6) cm s-1 bar-1, and was significantly greater in stage I sporangiophores than in stage IV sporangiophores. Contrary to previous reports, the sporangiophore is separated from the supporting mycelium by septa which prevent bulk volume flow between the two regions. The presence of a wall compartment between the cuticle and the plasma membrane results in anomalous osmosis during pressure clamp measurements. This behavior arises because of changes in solute concentration as water moves into or out of the wall compartment surrounding the sporangiophore. Theoretical analysis shows how the equations governing transient water flow are altered by the characteristics of the cell wall compartment.

Experiments on aerosol removal by high-pressure water spray

Energy Technology Data Exchange (ETDEWEB)

Corno, Ada del, E-mail: delcorno@rse-web.it [RSE, Power Generation Technologies and Materials Dept, via Rubattino 54, I-20134 Milano (Italy); Morandi, Sonia, E-mail: morandi@rse-web.it [RSE, Power Generation Technologies and Materials Dept, via Rubattino 54, I-20134 Milano (Italy); Parozzi, Flavio, E-mail: parozzi@rse-web.it [RSE, Power Generation Technologies and Materials Dept, via Rubattino 54, I-20134 Milano (Italy); Araneo, Lucio, E-mail: lucio.araneo@polimi.it [Politecnico di Milano, Department of Energy, via Lambruschini 4A, I-20156 Milano (Italy); CNR-IENI, via Cozzi 53, I-20125 Milano (Italy); Casella, Francesco, E-mail: francesco2.casella@mail.polimi.it [Politecnico di Milano, Department of Energy, via Lambruschini 4A, I-20156 Milano (Italy)

2017-01-15

Highlights: • Experimental research to measure the efficiency of high-pressure sprays in capturing aerosols if applied to a filtered containment venting system in case of severe accident. • Cloud of monodispersed SiO{sub 2} particles with sizes 0.5 or 1.0 μm and initial concentration in the range 2–90 mg/m{sup 3}. • Carried out in a chamber 0.5 × 1.0 m and 1.5 m high, with transparent walls equipped with a high pressure water spray with single nozzle. • Respect to low-pressure sprays, removal efficiency turned out significant: the half-life for 1 μm particles with a removal high-pressure spray system is orders of magnitude shorter than that with a low-pressure sprays system. - Abstract: An experimental research was managed in the framework of the PASSAM European Project to measure the efficiency of high-pressure sprays in capturing aerosols when applied to a filtered containment venting system in case of severe accident. The campaign was carried out in a purposely built facility composed by a scrubbing chamber 0.5 × 1.0 m and 1.5 m high, with transparent walls to permit the complete view of the aerosol removal process, where the aerosol was injected to form a cloud of specific particle concentration. The chamber was equipped with a high pressure water spray system with a single nozzle placed on its top. The test matrix consisted in the combination of water pressure injections, in the range 50–130 bar, on a cloud of monodispersed SiO{sub 2} particles with sizes 0.5 or 1.0 μm and initial concentration ranging between 2 and 99 mg/m{sup 3}. The spray was kept running for 2 min and the efficiency of the removal was evaluated, along the test time, using an optical particle sizer. With respect to low-pressure sprays, the removal efficiency turned out much more significant: the half-life for 1 μm particles with a removal high-pressure spray system is orders of magnitude shorter than that with a low-pressure spray system. The highest removal rate was

Experiments on aerosol removal by high-pressure water spray

International Nuclear Information System (INIS)

Corno, Ada del; Morandi, Sonia; Parozzi, Flavio; Araneo, Lucio; Casella, Francesco

2017-01-01

Highlights: • Experimental research to measure the efficiency of high-pressure sprays in capturing aerosols if applied to a filtered containment venting system in case of severe accident. • Cloud of monodispersed SiO_2 particles with sizes 0.5 or 1.0 μm and initial concentration in the range 2–90 mg/m"3. • Carried out in a chamber 0.5 × 1.0 m and 1.5 m high, with transparent walls equipped with a high pressure water spray with single nozzle. • Respect to low-pressure sprays, removal efficiency turned out significant: the half-life for 1 μm particles with a removal high-pressure spray system is orders of magnitude shorter than that with a low-pressure sprays system. - Abstract: An experimental research was managed in the framework of the PASSAM European Project to measure the efficiency of high-pressure sprays in capturing aerosols when applied to a filtered containment venting system in case of severe accident. The campaign was carried out in a purposely built facility composed by a scrubbing chamber 0.5 × 1.0 m and 1.5 m high, with transparent walls to permit the complete view of the aerosol removal process, where the aerosol was injected to form a cloud of specific particle concentration. The chamber was equipped with a high pressure water spray system with a single nozzle placed on its top. The test matrix consisted in the combination of water pressure injections, in the range 50–130 bar, on a cloud of monodispersed SiO_2 particles with sizes 0.5 or 1.0 μm and initial concentration ranging between 2 and 99 mg/m"3. The spray was kept running for 2 min and the efficiency of the removal was evaluated, along the test time, using an optical particle sizer. With respect to low-pressure sprays, the removal efficiency turned out much more significant: the half-life for 1 μm particles with a removal high-pressure spray system is orders of magnitude shorter than that with a low-pressure spray system. The highest removal rate was detected with 1 �

Cutting device for spent local power range monitor

International Nuclear Information System (INIS)

Watanabe, Shigeru; Tsuji, Teruaki.

1976-01-01

Purpose: To prevent radioactive contamination of the reactor water and simplify the operation of transferring the spent local power range monitor to a pool and preparation of the space for storing the monitor in the pool by arranging it such that the monitor is cut in the reactor water and that its cut end is closed in the reactor water. Constitution: A device for clamping a spent local power range monitor disposed within the reactor core and for cutting the monitor such that the content will not be exposed are supported by support means such that they are spaced apart in the vertical direction, and the support means is suspended in the reactor water such that it is movable relative to the monitor. Thus, there is no need to mount the cutter in any separately provided support means, and there is no possibility of exposure of the content since the cut end of the monitor is closed by the outer frame. It is therefore possible to prevent contamination of the reactor water. (Horiuchi, T.)

Sloshing of water in annular pressure-suppression pool of boiling water reactors under earthquake ground motions

International Nuclear Information System (INIS)

Aslam, M.; Godden, W.G.; Scalise, D.T.

1979-10-01

This report presents an analytical investigation of the sloshing response of water in annular-circular as well as simple-circular tanks under horizontal earthquake ground motions, and the results are verified with tests. This study was motivated because of the use of annular tanks for pressure-suppression pools in Boiling Water Reactors. Such a pressure-suppression pool would typically have 80 ft and 120 ft inside and outside diameters and a water depth of 20 ft. The analysis was based upon potential flow theory and a computer program was written to obtain time-history plots of sloshing displacements of water and the dynamic pressures. Tests were carried out on 1/80th and 1/15th scale models under sinusoidal as well as simulated earthquake ground motions. Tests and analytical results regarding the natural frequencies, surface water displacements, and dynamic pressures were compared and a good agreement was found for relatively small displacements. The computer program gave satisfactory results as long as the maximum water surface displacements were less than 30 in., which is roughly the value obtained under full intensity of El Centro earthquake

Logistics characterization for regional spent fuel repositories concept

International Nuclear Information System (INIS)

Joy, D.S.; Hudson, B.J.; Anthony, M.W.

1980-08-01

This report summarizes a study of logistics considerations for a four-region repository system for spent fuel disposal. The logistics considerations include: (1) yearly receipt and emplacement; (2) inventory; (3) away-from-reactor (AFR) storage; (4) nuclear capacity growth effects; (5) entire lifetime of reactors served by repository operations; (6) proportions of pressurized-water-reactor (PWR)/boiling-water-reactor (BWR) fuel; (7) proportions of rail and truck shipments; (8) shipping cask fleet requirements; (9) number of annual shipments; (10) mode (rail/truck) and cost of shipment; and (11) initial year for shipment to maintain full core reserve. The nation was divided into Northeast, North Central, Southern, and Western regions for evaluation purposes. Repository logistics were analyzed in each region based on three different capacity projections. For the Southern region, results for seven salt dome sites are presented. The Western region results cover four potential sites. The North Central and Northeastern regions results are not presented on a site specific basis. Conclusions are drawn based on the results. The methodology assumptions and references used in the logistics analysis are described for the convenience of the reader

Post-accident cooling capacity analysis of the AP1000 passive spent fuel pool cooling system

International Nuclear Information System (INIS)

Su Xia

2013-01-01

The passive design is used in AP1000 spent fuel pool cooling system. The decay heat of the spent fuel is removed by heating-boiling method, and makeup water is provided passively and continuously to ensure the safety of the spent fuel. Based on the analysis of the post-accident cooling capacity of the spent fuel cooling system, it is found that post-accident first 72-hour cooling under normal refueling condition and emergency full-core offload condition can be maintained by passive makeup from safety water source; 56 hours have to be waited under full core refueling condition to ensure the safety of the core and the spent fuel pool. Long-term cooling could be conducted through reserved safety interface. Makeup measure is available after accident and limited operation is needed. Makeup under control could maintain the spent fuel at sub-critical condition. Compared with traditional spent fuel pool cooling system design, the AP1000 design respond more effectively to LOCA accidents. (authors)

Improvement in fuel utilization in pressurized heavy water reactors due to increased heavy water purity

International Nuclear Information System (INIS)

Balakrishnan, M.R.

1991-01-01

This paper reports that in a pressurized heavy water reactor (PHWR), the reactivity of the reactor and, consequently, the discharge burnup of the fuel depend on the isotopic purity of the heavy water used in the reactor. The optimal purity of heavy water used in PHWRs, in turn, depends on the cost of fabricated uranium fuel and on the incremental cost incurred in improving the heavy water purity. The physics and economics aspects of the desirability of increasing the heavy water purity in PHWRs in India were first examined in 1978. With the cost data available at that time, it was found that improving the heavy water purity from 99.80% to 99.95% was economically attractive. The same problem is reinvestigated with current cost data. Even now, there is sufficient incentive to improve the isotopic purity of heavy water used in PHWRs. Admittedly, the economic advantage that can be derived depends on the cost of the fabricated fuel. Nevertheless, irrespective of the economics, there is also a fairly substantial saving in natural uranium. That the increase in the heavy water purity is to be maintained only in the low-pressure moderator system, and not in the high-pressure coolant system, makes the option of achieving higher fuel burnup with higher heavy water purity feasible

Spent nuclear fuel structural response when subject to an end impact accident

Energy Technology Data Exchange (ETDEWEB)

Tang, D.T.; Guttmann, J. [United States Nuclear Regulatory Commission, Rockville, MD (United States)]|[United States Nuclear Regulatory Commission, Washington, DC (United States); Koeppel, B.J.; Adkins, H.E.

2004-07-01

The US Nuclear Regulatory Commission (USNRC) is responsible for licensing spent fuel storage and transportation systems. A subset of this responsibility is to investigate and understand the structural performance of these systems. Studies have shown that the fuel rods of intact spent fuel assemblies with burn-ups up to 45 gigawatt days per metric ton of uranium (Gwd/MTU) are capable of resisting the normally expected impact loads subjected during drop accident conditions. However, effective cladding thickness for intact spent fuel assemblies with burn ups greater than 45 Gwd/MTU can be reduced due to corrosion. The capability of the fuel rod to withstand the expected loads encountered under normal and accident conditions may also be reduced, given degradation of the material properties under extended use, such as decrease in ductility. The USNRC and Pacific Northwest Laboratory (PNNL) performed computational studies to predict the structural response of spent nuclear fuel in a transport system that is subjected to a hypothetical regulatory impact accident, as defined in 10 CFR71.73. This study performs a structural analysis of a typical high burn up Pressurized Water Reactor (PWR) fuel assembly using the ANSYS {sup registered} ANSYS {sup registered} /LS- DYNA {sup registered} finite element analysis (FEA) code. The material properties used in the analyses were based on expert judgment and included uncertainties. Ongoing experimental programs will reduce the uncertainties. The current evaluations include the pins, spacer grids, and tie plates to assess possible cladding failure/rupture under hypothetical impact accident loading. This paper describes the USNRC and PNNL staff's analytical approach, provides details on the single pin model developed for this assessment, and presents the results.

Fundamentals of pressurized water reactors

International Nuclear Information System (INIS)

Murray, L.

1982-01-01

In many countries, the pressurized water reactor (PWR) is the most widely used, even though it requires enrichment of the uranium to about 3% in U-235 and the moderator-coolant must be maintained at a high pressure, about 2200 pounds per square inch. Our objective in this series of seven lectures is to describe the design and operating characteristics of the PWR system, discuss the reactor physics methods used to evaluate performance, examine the way fuel is consumed and produced, study the instrumentation system, review the physics measurements made during initial startup of the reactor, and outline the administrative aspects of starting up a reactor and operating it safely and effectively

Pyroprocessing oxide spent nuclear fuels for efficient disposal

International Nuclear Information System (INIS)

McPheeters, C.C.; Pierce, R.D.; Mulcahey, T.P.

1994-01-01

Pyrochemical processing as a means for conditioning spent nuclear fuels for disposal offers significant advantages over the direct disposal option. The advantages include reduction in high-level waste volume; conversion of most of the high-level waste to a low-level waste in which nearly all the transuranics (TRU) have been removed; and incorporation of the TRUs into a stable, highly radioactive waste form suitable for interim storage, ultimate destruction, or repository disposal. The lithium process has been under development at Argonne National Laboratory for use in pyrochemical conditioning of spent fuel for disposal. All of the process steps have been demonstrated in small-scale (0.5-kg simulated spent fuel) experiments. Engineering-scale (20-kg simulated spent fuel) demonstration of the process is underway, and small-scale experiments have been conducted with actual spent fuel from a light water reactor (LWR). The lithium process is simple, operates at relatively low temperatures, and can achieve high decontamination factors for the TRU elements. Ordinary materials, such as carbon steel, can be used for process containment

46 CFR 52.01-110 - Water-level indicators, water columns, gauge-glass connections, gauge cocks, and pressure gauges...

Science.gov (United States)

2010-10-01

... § 52.01-110 Water-level indicators, water columns, gauge-glass connections, gauge cocks, and pressure... 46 Shipping 2 2010-10-01 2010-10-01 false Water-level indicators, water columns, gauge-glass connections, gauge cocks, and pressure gauges (modifies PG-60). 52.01-110 Section 52.01-110 Shipping COAST...

Comprehensive studies on regulatory issues of spent fuel pools

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

An existence of safety issues in the spent fuel pool (SFP) was recognized by the nuclear accident at the Fukushima Daiichi Nuclear Power Station, and many reports on the accident describe needs of countermeasures for SFP under sever accidents. For research planning, thermal hydraulic behaviors of SFP and possibility of occurrence of re-criticality conditions in SFP were studied by computational approaches. In the studies on thermal hydraulic behaviors, possibilities of adiabatic conditions in a spent fuel bundle were identified because natural circulation cooling of air could be terminated due to flow path blockage by pool water and steam cooling could be terminated due to reduction of pool water evaporation originated from cold water injection by emergency water supply. In the re-criticality study, in the case of the un-borated lack, it was shown that the neutron multiplication factor became larger than unity when the difference of water levels inside and outside the channel box larger than some values. (author)

Spent fuel dry storage technology development: thermal evaluation of three adjacent drywells (each containing a 0.6 kW PWR spent fuel assembly)

International Nuclear Information System (INIS)

Unterzuber, R.; Hanson, J.P.

1981-09-01

A spent fuel Adjacent Drywell Test was conducted at the Engine-Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site utilizing three nearly identical pressurized water reactor spent fuel assemblies each having a decay heat level of approximately 0.6 kW. Each fuel assembly was encapsulated inside the E-MAD Hot Bay and placed in an instrumented near-surface drywell storage cell for thermal testing. Each fuel assembly was sealed inside a 14-in. diam, 168-in.-long stainless steel canister and attached to a concrete-filled, 20-in.-diam, 34-in.-long, shield plug. The canister assembly was then placed in a carbon steel drywell liner which had been grouted into a hole drilled in the soil adjacent to E-MAD. The three drywells were located 25 feet apart in a linear array. Thermocouples, provided to measure canister, liner and soil temperatures, were inserted into tubes on the outside of the canister and drywell liner and were attached to plastic pipes which were grouted into holes in the soil. Temperatures from the three drywells and the adjacent soil were recorded throughout the Adjacent Drywell Test. Drywell thermal data showed virtually no thermal interaction between adjacent drywells. However, peak temperatures reached by the three drywells did show a fairly significant difference. Peak canister and drywell liner temperatures were reached in August 1981 for all three drywells. The two previously unused drywells responded similarly with peak canister and liner temperatures reaching 199 0 F and 158 0 F, respectively. Comparable peak temperatures for the third drywell which had previously contained spent fuel for nearly 21 months prior to the Adjacent Drywell Test reached 210 0 F for the canister and 169 0 F for the drywell liner. This difference is attributed to a decrease in soil thermal conductivity caused by the dryout of soil around the drywell used for previous spent fuel testing

Corrosion of aluminium-clad spent fuel at RA research reactor

International Nuclear Information System (INIS)

Pesic, M.; Maksin, T.; Dobrijevic, R.; Idjakovic, Z.

2003-01-01

Almost 95% of all spent fuel elements of the RA research reactor in the Vinca Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, are stored in 30 aluminium barrels and about 300 stainless steel channel-holders in the temporary spent fuel storage water pool. The first activities of sludge and water samples, taken from the pool, were measured in 1996-1997 and were followed by analysis of chemical composition of samples. Visual inspections of fuel elements in some stainless steel tubes and of the fuel channels stored in the reactor core have shown that some deposits cover aluminium cladding. Stains and surface discoloration are noted on many of the spent fuel elements that were examined visually during the core unloading and inspections carried out in 1979 - 1984. Some of water samples, taken from pool, about a 150 stainless steel tubes and 16 barrels have shown very high 137-Cs activity compared to low activity measured in pool water. It was concluded that aluminium cladding of the fuel elements was penetrated due to corrosion process. Study on influence of water corrosion processes in the RA reactor storage pool was started within the framework of the IAEA CRP 'Corrosion of Research Reactor Aluminium-Clad Spent Fuel in Water' in 2002. The first test rack with various aluminium and stainless steel coupons, supplied by the IAEA, was immersed in the pool already in 1996. New racks were immersed in 2002 and 2003. The rack immersed in 1996 was taken out from the pool in 2002 and the rack immersed in 2002 was taken out in 2003. Results of the examination of these racks, carried out according to the strategy and the protocol, proposed by the IAEA, are described in this paper. (author)

Radiochemical analyses of several spent fuel Approved Testing Materials

International Nuclear Information System (INIS)

Guenther, R.J.; Blahnik, D.E.; Wildung, N.J.

1994-09-01

Radiochemical characterization data are described for UO 2 and UO 2 plus 3 wt% Gd 2 O 3 commercial spent nuclear fuel taken from a series of Approved Testing Materials (ATMs). These full-length nuclear fuel rods include MLA091 of ATM-103, MKP070 of ATM-104, NBD095 and NBD131 of ATM-106, and ADN0206 of ATM-108. ATMs 103, 104, and 106 were all irradiated in the Calvert Cliffs Nuclear Power Plant (Reactor No.1), a pressurized-water reactor that used fuel fabricated by Combustion Engineering. ATM-108 was part of the same fuel bundle designed as ATM-105 and came from boiling-water reactor fuel fabricated by General Electric and irradiated in the Cooper Nuclear Power Plant. Rod average burnups and expected fission gas releases ranged from 2,400 to 3,700 GJ/kgM. (25 to 40 Mwd/kgM) and from less than 1% to greater than 10%, respectively, depending on the specific ATM. The radiochemical analyses included uranium and plutonium isotopes in the fuel, selected fission products in the fuel, fuel burnup, cesium and iodine on the inner surfaces of the cladding, 14 C in the fuel and cladding, and analyses of the gases released to the rod plenum. Supporting examinations such as fuel rod design and material descriptions, power histories, and gamma scans used for sectioning diagrams are also included. These ATMs were examined as part of the Materials Characterization Center Program conducted at Pacific Northwest Laboratory provide a source of well-characterized spent fuel for testing in support of the US Department of Energy Office of Civilian Radioactive Waste Management Program

Spent LWR fuel encapsulation and dry storage demonstration

International Nuclear Information System (INIS)

Bahorich, R.J.; Durrill, D.C.; Cross, T.E.; Unterzuber, R.

1980-01-01

In 1977 the Spent Fuel Handling and Packaging Program (SFHPP) was initiated by the Department of Energy to develop and test the capability to satisfactorily encapsulate typical spent fuel assemblies from commercial light-water nuclear power plants and to establish the suitability of one or more surface and near surface concepts for the interim dry storage of the encapsulated spent fuel assemblies. The E-MAD Facility at the Nevada Test Site, which is operated for the Department of Energy by the Advanced Energy Systems Division (AESD) of the Westinghouse Electric Corporation, was chosen as the location for this demonstration because of its extensive existing capabilities for handling highly radioactive components and because of the desirable site characteristics for the proposed storage concepts. This paper describes the remote operations related to the process steps of handling, encapsulating and subsequent dry storage of spent fuel in support of the Demonstration Program

Conceptual design of a system for detecting national diversion of LWR spent fuel

International Nuclear Information System (INIS)

Holmes, J.P.

1978-09-01

A conceptual design for detecting the national diversion of light water reactor spent fuel in water basin storage or in transit between facilities is described. This is the third in a series of reports dealing with this topic. The first report provides the spent fuel facilities and operations baseline description; the second report discusses cost and performance tradeoffs for three inspection and surveillance concepts for the detection of a national diversion of spent fuel. The conceptual design presented herein will provide a basis for future feasibility investigations and tradeoff analyses of hardware configurations and inspection options

Evaluating the Laplace pressure of water nanodroplets from simulations

Science.gov (United States)

Malek, Shahrazad M. A.; Sciortino, Francesco; Poole, Peter H.; Saika-Voivod, Ivan

2018-04-01

We calculate the components of the microscopic pressure tensor as a function of radial distance r from the centre of a spherical water droplet, modelled using the TIP4P/2005 potential. To do so, we modify a coarse-graining method for calculating the microscopic pressure (Ikeshoji et al 2003 Mol. Simul. 29 101) in order to apply it to a rigid molecular model of water. As test cases, we study nanodroplets ranging in size from 776 to 2880 molecules at 220 K. Beneath a surface region comprising approximately two molecular layers, the pressure tensor becomes approximately isotropic and constant with r. We find that the dependence of the pressure on droplet radius is that expected from the Young-Laplace equation, despite the small size of the droplets.

Peak cladding temperature in a spent fuel storage or transportation cask

International Nuclear Information System (INIS)

Li, J.; Murakami, H.; Liu, Y.; Gomez, P.E.A.; Gudipati, M.; Greiner, M.

2007-01-01

From reactor discharge to eventual disposition, spent nuclear fuel assemblies from a commercial light water reactor are typically exposed to a variety of environments under which the peak cladding temperature (PCT) is an important parameter that can affect the characteristics and behavior of the cladding and, thus, the functions of the spent fuel during storage, transportation, and disposal. Three models have been identified to calculate the peak cladding temperature of spent fuel assemblies in a storage or transportation cask: a coupled effective thermal conductivity and edge conductance model developed by Manteufel and Todreas, an effective thermal conductivity model developed by Bahney and Lotz, and a computational fluid dynamics model. These models were used to estimate the PCT for spent fuel assemblies for light water reactors under helium, nitrogen, and vacuum environments with varying decay heat loads and temperature boundary conditions. The results show that the vacuum environment is more challening than the other gas environments in that the PCT limit is exceeded at a lower boundary temperature for a given decay heat load of the spent fuel assembly. This paper will highlight the PCT calculations, including a comparison of the PCTs obtained by different models.

LINEAR KERNEL SUPPORT VECTOR MACHINES FOR MODELING PORE-WATER PRESSURE RESPONSES

Directory of Open Access Journals (Sweden)

KHAMARUZAMAN W. YUSOF

2017-08-01

Full Text Available Pore-water pressure responses are vital in many aspects of slope management, design and monitoring. Its measurement however, is difficult, expensive and time consuming. Studies on its predictions are lacking. Support vector machines with linear kernel was used here to predict the responses of pore-water pressure to rainfall. Pore-water pressure response data was collected from slope instrumentation program. Support vector machine meta-parameter calibration and model development was carried out using grid search and k-fold cross validation. The mean square error for the model on scaled test data is 0.0015 and the coefficient of determination is 0.9321. Although pore-water pressure response to rainfall is a complex nonlinear process, the use of linear kernel support vector machine can be employed where high accuracy can be sacrificed for computational ease and time.

Microwave measurements of water vapor partial pressure at high temperatures

International Nuclear Information System (INIS)

Latorre, V.R.

1991-01-01

One of the desired parameters in the Yucca Mountain Project is the capillary pressure of the rock comprising the repository. This parameter is related to the partial pressure of water vapor in the air when in equilibrium with the rock mass. Although there are a number of devices that will measure the relative humidity (directly related to the water vapor partial pressure), they generally will fail at temperatures on the order of 150C. Since thee author has observed borehole temperatures considerably in excess of this value in G-Tunnel at the Nevada Test Site (NTS), a different scheme is required to obtain the desired partial pressure data at higher temperatures. This chapter presents a microwave technique that has been developed to measure water vapor partial pressure in boreholes at temperatures up to 250C. The heart of the system is a microwave coaxial resonator whose resonant frequency is inversely proportional to the square root of the real part of the complex dielectric constant of the medium (air) filling the resonator. The real part of the dielectric constant of air is approximately equal to the square of the refractive index which, in turn, is proportional to the partial pressure of the water vapor in the air. Thus, a microwave resonant cavity can be used to measure changes in the relative humidity or partial pressure of water vapor in the air. Since this type of device is constructed of metal, it is able to withstand very high temperatures. The actual limitation is the temperature limit of the dielectric material in the cable connecting the resonator to its driving and monitoring equipment-an automatic network analyzer in our case. In the following sections, the theory of operation, design, construction, calibration and installation of the microwave diagnostics system is presented. The results and conclusions are also presented, along with suggestions for future work

Loss of spent fuel pool cooling PRA: Model and results

International Nuclear Information System (INIS)

Siu, N.; Khericha, S.; Conroy, S.; Beck, S.; Blackman, H.

1996-09-01

This letter report documents models for quantifying the likelihood of loss of spent fuel pool cooling; models for identifying post-boiling scenarios that lead to core damage; qualitative and quantitative results generated for a selected plant that account for plant design and operational practices; a comparison of these results and those generated from earlier studies; and a review of available data on spent fuel pool accidents. The results of this study show that for a representative two-unit boiling water reactor, the annual probability of spent fuel pool boiling is 5 x 10 -5 and the annual probability of flooding associated with loss of spent fuel pool cooling scenarios is 1 x 10 -3 . Qualitative arguments are provided to show that the likelihood of core damage due to spent fuel pool boiling accidents is low for most US commercial nuclear power plants. It is also shown that, depending on the design characteristics of a given plant, the likelihood of either: (a) core damage due to spent fuel pool-associated flooding, or (b) spent fuel damage due to pool dryout, may not be negligible

Frictional pressure drop of high pressure steam-water two-phase flow in internally helical ribbed tubes

International Nuclear Information System (INIS)

Tingkuan, C.; Xuanzheng, C.

1987-01-01

It is well known that the internally helical ribbed tubes are effective in suppressing the dry-out in boiling tubes at high pressures, so they are widely used as furnace water wall tubes in modern large steam power boilers. Design of the boilers requires the data on frictional pressure drop characteristics of the ribbed tubes, but they are not sufficient now. This paper describes the experimental results on the adiabatic frictional pressure drop in both horizontal ribbed tubes with measured mean inside diameter of 11.69 mm and 35.42 mm at high pressure from 10 to 21 MPa, mass flow rate from 350 to 3800 kg/m/sup 2/s and steam quality from 0 to 1 in our high pressure electrically heated water loop. Simultaneously, both smooth tubes under the same conditions for comparison. Based on the tests the correlation for determining the frictional pressure drop of internally ribbed tubes are proposed

Design of a Prototype Differential Die‐Away Instrument Proposed for Swedish Spent Nuclear Fuel Characterization

Energy Technology Data Exchange (ETDEWEB)

Martinik, Tomas, E-mail: tomas.martinik@physics.uu.se [Department of Physics and Astronomy, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Henzl, Vladimir [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Grape, Sophie; Jansson, Peter [Department of Physics and Astronomy, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); Swinhoe, Martyn T.; Goodsell, Alison V. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Tobin, Stephen J. [Department of Physics and Astronomy, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Swedish Nuclear Fuel and Waste Management Company, Blekholmstorget 30, Box 250, SE-101 24 Stockholm (Sweden)

2016-06-11

As part of the United States (US) Department of Energy's Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project, the traditional Differential Die-Away (DDA) method that was originally developed for waste drum assay has been investigated and modified to provide a novel application to characterize or verify spent nuclear fuel (SNF). Following the promising, yet largely theoretical and simulation based, research of physics aspects of the DDA technique applied to SNF assay during the early stages of the NGSI-SF project, the most recent effort has been focused on the practical aspects of developing the first fully functional and deployable DDA prototype instrument for spent fuel. As a result of the collaboration among US research institutions and Sweden, the opportunity to test the newly proposed instrument's performance with commercial grade SNF at the Swedish Interim Storage Facility (Clab) emerged. Therefore the design of this instrument prototype has to accommodate the requirements of the Swedish regulator as well as specific engineering constrains given by the unique industrial environment. Within this paper, we identify key components of the DDA based instrument and we present methodology for evaluation and the results of a selection of the most relevant design parameters in order to optimize the performance for a given application, i.e. test-deployment, including assay of 50 preselected spent nuclear fuel assemblies of both pressurized (PWR) as well as boiling (BWR) water reactor type.

Oxidizing dissolution of spent MOX47 fuel subjected to water radiolysis: Solution chemistry and surface characterization by Raman spectroscopy

Energy Technology Data Exchange (ETDEWEB)

Jegou, C., E-mail: christophe.jegou@cea.f [Commissariat a l' Energie Atomique (CEA), Marcoule Reasearch Center, B.P. 17171, F-30207 Bagnols-sur-Ceze Cedex (France); Caraballo, R.; De Bonfils, J.; Broudic, V.; Peuget, S. [Commissariat a l' Energie Atomique (CEA), Marcoule Reasearch Center, B.P. 17171, F-30207 Bagnols-sur-Ceze Cedex (France); Vercouter, T. [Commissariat a l' Energie Atomique (CEA), Saclay Reasearch Center, B.P. 11, F-91191 Gif-sur-Yvette Cedex (France); Roudil, D. [Commissariat a l' Energie Atomique (CEA), Marcoule Reasearch Center, B.P. 17171, F-30207 Bagnols-sur-Ceze Cedex (France)

2010-04-01

The mechanisms of oxidizing dissolution of spent MOX fuel (MIMAS TU2 (registered) ) subjected to water radiolysis were investigated experimentally by leaching spent MOX47 fuel samples in pure water at 25 deg. C under different oxidizing conditions (with and without external gamma irradiation); the leached surfaces were characterized by Raman spectroscopy. The highly oxidizing conditions resulting from external gamma irradiation significantly increased the concentration of plutonium (Pu(V)) and uranium (U(VI)) compared with a benchmark experiment (without external irradiation). The oxidation behavior of the plutonium-enriched aggregates differed significantly from that of the UO{sub 2} matrix after several months of leaching in water under gamma irradiation. The plutonium in the aggregates appears to limit fuel oxidation. The only secondary phases formed and identified to date by Raman spectroscopy are uranium peroxides that generally precipitate on the surface of the UO{sub 2} grains. Concerning the behavior of plutonium, solution analysis results appear to be compatible with a conventional explanation based on an equilibrium with a Pu(OH){sub 4(am)} phase. The fission product release - considered as a general indicator of matrix alteration - from MOX47 fuel also increases under external gamma irradiation and a change in the leaching mode is observed. Diffusive leaching was clearly identified, coinciding with the rapid onset of steady-state actinide concentrations in the bulk solution.

The future 700 MWe pressurized heavy water reactor

International Nuclear Information System (INIS)

Bhardwaj, S.A.

2006-01-01

The design of a 700 MWe pressurized heavy water reactor has been developed. The design is based on the twin 540 MWe reactors at Tarapur of which the first unit has been made critical in less than 5 years from construction commencement. In the 700 MWe design boiling of the coolant, to a limited extent, has been allowed near the channel exit. While making the plant layout more compact, emphasis has been on constructability. Saving in capital cost of about 15%, over the present units, is expected. The paper describes salient design features of 700 MWe pressurized heavy water reactor

Assessment of Analytic Water hammer Pressure Model of FAI/08-70

International Nuclear Information System (INIS)

Park, Ju Yeop; Yoo, Seung Hun; Seul, Kwang-Won

2016-01-01

In evaluating water hammer effect on the safety related systems, methods developed by the US utility are likely to be adopted in Korea. For example, the US utility developed specific methods to evaluate pressure and loading transient on piping due to water hammer as in FAI/08-70. The methods of FAI/08-70 would be applied in Korea when any regulatory request on the evaluation of water hammer effect due to the non-condensable gas accumulation in the safety related systems. Specifically, FAI/08-70 gives an analytic model which can be used to analyze the maximum transient pressure and maximum transient loading on the piping of the safety-related systems due to the non-condensable induced water hammer effect. Therefore, it seems to be meaningful to review the FAI/08-70 methods and attempt to apply the methods to a specific case to see if they really give reasonable estimate before the application of FAI/08-70 methods to domestic nuclear power plants. In the present study, analytic water hammer pressure model of FAI/08-70 is reviewed in detail and the model is applied to the specific experiment of FAI/08-70 to see if the analytic water hammer pressure model really gives reasonable estimate of the peak water hammer pressure. Specifically, we assess the experiment 52A of FAI/08-70 which adopts flushed initial condition with a short rising piping length and a high level piping length of 51inch. The calculated analytic water hammer pressure peak shows a close agreement with the measured experimental data of 52A. Unfortunately, however, the theoretical value is a little bit less than that of the experimental value. This implies the analytic model of FAI/08-70 is not conservative

Assessment of Analytic Water hammer Pressure Model of FAI/08-70

Energy Technology Data Exchange (ETDEWEB)

Park, Ju Yeop; Yoo, Seung Hun; Seul, Kwang-Won [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2016-10-15

In evaluating water hammer effect on the safety related systems, methods developed by the US utility are likely to be adopted in Korea. For example, the US utility developed specific methods to evaluate pressure and loading transient on piping due to water hammer as in FAI/08-70. The methods of FAI/08-70 would be applied in Korea when any regulatory request on the evaluation of water hammer effect due to the non-condensable gas accumulation in the safety related systems. Specifically, FAI/08-70 gives an analytic model which can be used to analyze the maximum transient pressure and maximum transient loading on the piping of the safety-related systems due to the non-condensable induced water hammer effect. Therefore, it seems to be meaningful to review the FAI/08-70 methods and attempt to apply the methods to a specific case to see if they really give reasonable estimate before the application of FAI/08-70 methods to domestic nuclear power plants. In the present study, analytic water hammer pressure model of FAI/08-70 is reviewed in detail and the model is applied to the specific experiment of FAI/08-70 to see if the analytic water hammer pressure model really gives reasonable estimate of the peak water hammer pressure. Specifically, we assess the experiment 52A of FAI/08-70 which adopts flushed initial condition with a short rising piping length and a high level piping length of 51inch. The calculated analytic water hammer pressure peak shows a close agreement with the measured experimental data of 52A. Unfortunately, however, the theoretical value is a little bit less than that of the experimental value. This implies the analytic model of FAI/08-70 is not conservative.

1. The application of PIE techniques to the study of the corrosion of spent oxide fuel in deep-rock groundwaters. 2. Spent fuel degradation

International Nuclear Information System (INIS)

Forsyth, R.S.

1991-01-01

During the autumn of 1990, papers summarizing work performed at Studsvik as part of the SKB research programme designed to study the corrosion behaviour of spent nuclear fuel in deep-rock groundwater were presented at two scientific meetings: The first paper presents results and observations of the study of the corrosion of spent oxide fuel in deep-rock ground-waters. The PIE techniques were applied to the detailed study of spent fuel both before and after water contact. The second paper represents an up-dated reporting of results obtained in the Swedish programme relevant to preferential dissolution effects, including interim results from recently stored experiments specifically designed to study possible correlations between corrosion behaviour and fuel properties conditioned by burnup and/or local power variations. Recent observations during the search for corrosion sites in fuel exposed to corrosion for about 4 years are also presented. (KAE)

Immobilization of radioactive waste sludge from spent fuel storage pool

International Nuclear Information System (INIS)

Pavlovic, R.; Plecas, I.

1998-01-01

In the last forty years, in FR Yugoslavia, as result of the research reactors' operation and radionuclides application in medicine, industry and agriculture, radioactive waste materials of the different categories and various levels of specific activities were generated. As a temporary solution, these radioactive waste materials are stored in the two hanger type interim storages for solid waste and some type of liquid waste packed in plastic barrels, and one of three stainless steal underground containers for other types of liquid waste. Spent fuel elements from nuclear reactors in the Vinca Institute have been temporary stored in water filled storage pool. Due to the fact that the water in the spent fuel elements storage pool have not been purified for a long time, all metallic components submerged in the water have been hardly corroded and significant amount of the sludge has been settled on the bottom of the pool. As a first step in improving spent fuel elements storage conditions and slowing down corrosion in the storage spent fuel elements pool we have decided to remove the sludge from the bottom of the pool. Although not high, but slightly radioactive, this sludge had to be treated as radioactive waste material. Some aspects of immobilisation, conditioning and storage of this sludge are presented in this paper. (author

Corrosion of spent Advanced Test Reactor fuel

International Nuclear Information System (INIS)

Lundberg, L.B.; Croson, M.L.

1994-01-01

The results of a study of the condition of spent nuclear fuel elements from the Advanced Test Reactor (ATR) currently being stored underwater at the Idaho National Engineering Laboratory (INEL) are presented. This study was motivated by a need to estimate the corrosion behavior of dried, spent ATR fuel elements during dry storage for periods up to 50 years. The study indicated that the condition of spent ATR fuel elements currently stored underwater at the INEL is not very well known. Based on the limited data and observed corrosion behavior in the reactor and in underwater storage, it was concluded that many of the fuel elements currently stored under water in the facility called ICPP-603 FSF are in a degraded condition, and it is probable that many have breached cladding. The anticipated dehydration behavior of corroded spent ATR fuel elements was also studied, and a list of issues to be addressed by fuel element characterization before and after forced drying of the fuel elements and during dry storage is presented

Development of Extended Period Pressure-Dependent Demand Water Distribution Models

Energy Technology Data Exchange (ETDEWEB)

Judi, David R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mcpherson, Timothy N. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-03-20

Los Alamos National Laboratory (LANL) has used modeling and simulation of water distribution systems for N-1 contingency analyses to assess criticality of water system assets. Critical components considered in these analyses include pumps, tanks, and supply sources, in addition to critical pipes or aqueducts. A contingency represents the complete removal of the asset from system operation. For each contingency, an extended period simulation (EPS) is run using EPANET. An EPS simulates water system behavior over a time period, typically at least 24 hours. It assesses the ability of a system to respond and recover from asset disruption through distributed storage in tanks throughout the system. Contingencies of concern are identified as those in which some portion of the water system has unmet delivery requirements. A delivery requirement is defined as an aggregation of water demands within a service area, similar to an electric power demand. The metric used to identify areas of unmet delivery requirement in these studies is a pressure threshold of 15 pounds per square inch (psi). This pressure threshold is used because it is below the required pressure for fire protection. Any location in the model with pressure that drops below this threshold at any time during an EPS is considered to have unmet service requirements and is used to determine cascading consequences. The outage area for a contingency is the aggregation of all service areas with a pressure below the threshold at any time during the EPS.

Methods for expanding the capacity of spent fuel storage facilities

International Nuclear Information System (INIS)

1990-06-01

At the beginning of 1989 more than 55,000 metric tonnes of heavy metal (MTHM) of spent Light Water Reactor (LWR) and Heavy Water Reactor (HWR) fuel had been discharged worldwide from nuclear power plants. Only a small fraction of this fuel has been reprocessed. The majority of the spent fuel assemblies are currently held at-reactor (AR) or away-from-reactor (AFR) in storage awaiting either chemical processing or final disposal depending on the fuel concept chosen by individual countries. Studies made by NEA and IAEA have projected that annual spent fuel arising will reach about 10,000 t HM in the year 2000 and cumulative arising will be more than 200,000 t HM. Taking into account the large quantity of spent fuel discharged from NPP and that the first demonstrations of the direct disposal of spent fuel or HLW are expected only after the year 2020, long-term storage will be the primary option for management of spent fuel until well into the next century. There are several options to expand storage capacity: (1) to construct new away-from-reactor storage facilities, (2) to transport spent fuel from a full at-reactor pool to another site for storage in a pool that has sufficient space to accommodate it, (3) to expand the capacity of existing AR pools by using compact racks, double-tierce, rod consolidation and by increasing the dimensions of existing pools. The purpose of the meeting was: to exchange new information on the international level on the subject connected with the expansion of storage capacities for spent fuel; to elaborate the state-of-the-art of this problem; to define the most important areas for future activity; on the basis of the above information to give recommendations to potential users for selection and application of the most suitable methods for expanding spent fuel facilities taking into account the relevant country's conditions. Refs, figs and tabs

Corrosion surveillance in spent fuel storage pools

International Nuclear Information System (INIS)

Howell, J.P.

1996-01-01

In mid-1991, corrosion of aluminum-clad spent nuclear fuel was observed in the light-water filled basins at the Savannah River site. A corrosion surveillance program was initiated in the P, K, L-Reactor basins and in the Receiving Basin for Offsite Fuels (RBOF). This program verified the aggressive nature of the pitting corrosion and provided recommendations for changes in basin operations to permit extended longer term interim storage. The changes were implemented during 1994--1996 and have resulted in significantly improved basin water quality with conductivity in the 1--3 microS/cm range. Under these improved conditions, no new pitting has been observed over the last three years. This paper describes the corrosion surveillance program at SRS and what has been learned about the corrosion of aluminum-clad in spent fuel storage pools

Ultra-high pressure water jetting for coating removal and surface preparation

Science.gov (United States)

Johnson, Spencer T.

1995-01-01

This paper shall examine the basics of water technology with particular attention paid to systems currently in use and some select new applications. By providing an overview of commercially available water jet systems in the context of recent case histories, potential users may evaluate the process for future applications. With the on going introduction of regulations prohibiting the use of chemical paint strippers, manual scrapping and dry abrasive media blasting, the need for an environmentally compliant coating removal process has been mandated. Water jet cleaning has been a traditional part of many industrial processed for year, although it has only been in the last few years that reliable pumping equipment capable of ultra-high pressure operation have become available. With the advent of water jet pumping equipment capable of sustaining pressures in excess of 36,000 psi. there has been shift away from lower pressure, high water volume systems. One of the major factors in driving industry to seek higher pressures is the ability to offer higher productivity rates while lowering the quantity of water used and subsequently reprocessed. Among benefits of the trend toward higher pressure/lower volume systems is the corresponding reduction in water jet reaction forces making hand held water jetting practical and safe. Other unique applications made possible by these new generation pumping systems include the use of alternative fluids including liquid ammonia for specialized and hazardous material removal applications. A review of the equipment used and the required modifications will be presented along with the conclusions reached reached during this test program.

Electrochemical noise measurements under pressurized water reactor conditions

International Nuclear Information System (INIS)

Van Nieuwenhove, R.

2000-01-01

Electrochemical potential noise measurements on sensitized stainless steel pressure tubes under pressurized water reactor (PWR) conditions were performed for the first time. Very short potential spikes, believed to be associated to crack initiation events, were detected when stressing the sample above the yield strength and increased in magnitude until the sample broke. Sudden increases of plastic deformation, as induced by an increased tube pressure, resulted in slower, high-amplitude potential transients, often accompanied by a reduction in noise level

Evaluation of Pad 18 Spent Mercury Gold Trap Stainless Steel Container Failure

International Nuclear Information System (INIS)

Skidmore, E.

2016-01-01

Failure of the Pad 18 spent mercury gold trap stainless steel waste container is principally attributed to corrosion induced by degradation of plasticized polyvinyl chloride (pPVC) waste packaging material. Dehydrochlorination of pPVC polymer by thermal and/or radiolytic degradation is well-known to evolve HCl gas, which is highly corrosive to stainless steel and other metals in the presence of moisture. Degradation of the pPVC packaging material was likely caused by radiolysis in the presence of tritium gas within the waste container, though other degradation mechanisms (aging, thermo-oxidation, plasticizer migration) over 30 years storage may have contributed. Corrosion was also likely enhanced by the crevice in the container weld design, and may have been enhanced by the presence of tritiated water. Similar non-failed spent mercury gold trap waste containers did not show radiographic evidence of plastic packaging or trapped free liquid within the container. Therefore, those containers are not expected to exhibit similar failures. Halogenated polymers such as pPVC subject to degradation can evolve halide gases such as HCl, which is corrosive in the presence of moisture and can generate pressure in sealed systems.

Evaluation of Pad 18 Spent Mercury Gold Trap Stainless Steel Container Failure

Energy Technology Data Exchange (ETDEWEB)

Skidmore, E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-08-03

Failure of the Pad 18 spent mercury gold trap stainless steel waste container is principally attributed to corrosion induced by degradation of plasticized polyvinyl chloride (pPVC) waste packaging material. Dehydrochlorination of pPVC polymer by thermal and/or radiolytic degradation is well-known to evolve HCl gas, which is highly corrosive to stainless steel and other metals in the presence of moisture. Degradation of the pPVC packaging material was likely caused by radiolysis in the presence of tritium gas within the waste container, though other degradation mechanisms (aging, thermo-oxidation, plasticizer migration) over 30 years storage may have contributed. Corrosion was also likely enhanced by the crevice in the container weld design, and may have been enhanced by the presence of tritiated water. Similar non-failed spent mercury gold trap waste containers did not show radiographic evidence of plastic packaging or trapped free liquid within the container. Therefore, those containers are not expected to exhibit similar failures. Halogenated polymers such as pPVC subject to degradation can evolve halide gases such as HCl, which is corrosive in the presence of moisture and can generate pressure in sealed systems.

Characteristics of fuel crud and its impact on storage, handling, and shipment of spent fuel

International Nuclear Information System (INIS)

Hazelton, R.F.

1987-09-01

Corrosion products, called ''crud,'' form on out-of-reactor surfaces of nuclear reactor systems and are transported by reactor coolant to the core, where they deposit on external fuel-rod cladding surfaces and are activated by nuclear reactions. After discharge of spent fuel from a reactor, spallation of radioactive crud from the fuel rods could impact wet or dry storage operations, handling (including rod consolidation), and shipping. It is the purpose of this report to review earlier (1970s) and more recent (1980s) literature relating to crud, its characteristics, and any impact it has had on actual operations. Crud characteristics vary from reactor type to reactor type, reactor to reactor, fuel assembly to fuel assembly in a reactor, circumferentially and axially in an assembly, and from cycle to cycle for a specific facility. To characterize crud of pressurized-water (PWRs) and boiling-water reactors (BWRs), published information was reviewed on appearance, chemical composition, areal density and thickness, structure, adhesive strength, particle size, and radioactivity. Information was also collected on experience with crud during spent fuel wet storage, rod consolidation, transportation, and dry storage. From experience with wet storage, rod consolidation, transportation, and dry storage, it appears crud spallation can be managed effectively, posing no significant radiological problems. 44 refs., 11 figs

Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

International Nuclear Information System (INIS)

Krishnan, S.; Bhasin, V.; Mahajan, S.C.

1997-01-01

Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300 degrees C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered

Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

Energy Technology Data Exchange (ETDEWEB)

Krishnan, S.; Bhasin, V.; Mahajan, S.C. [Bhabha Atomic Research Centre, Bombay (India)] [and others

1997-04-01

Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300{degrees}C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered.

Low internal pressure in femtoliter water capillary bridges reduces evaporation rates.

Science.gov (United States)

Cho, Kun; Hwang, In Gyu; Kim, Yeseul; Lim, Su Jin; Lim, Jun; Kim, Joon Heon; Gim, Bopil; Weon, Byung Mook

2016-03-01

Capillary bridges are usually formed by a small liquid volume in a confined space between two solid surfaces. They can have a lower internal pressure than the surrounding pressure for volumes of the order of femtoliters. Femtoliter capillary bridges with relatively rapid evaporation rates are difficult to explore experimentally. To understand in detail the evaporation of femtoliter capillary bridges, we present a feasible experimental method to directly visualize how water bridges evaporate between a microsphere and a flat substrate in still air using transmission X-ray microscopy. Precise measurements of evaporation rates for water bridges show that lower water pressure than surrounding pressure can significantly decrease evaporation through the suppression of vapor diffusion. This finding provides insight into the evaporation of ultrasmall capillary bridges.

Low internal pressure in femtoliter water capillary bridges reduces evaporation rates

Science.gov (United States)

Cho, Kun; Hwang, In Gyu; Kim, Yeseul; Lim, Su Jin; Lim, Jun; Kim, Joon Heon; Gim, Bopil; Weon, Byung Mook

2016-01-01

Capillary bridges are usually formed by a small liquid volume in a confined space between two solid surfaces. They can have a lower internal pressure than the surrounding pressure for volumes of the order of femtoliters. Femtoliter capillary bridges with relatively rapid evaporation rates are difficult to explore experimentally. To understand in detail the evaporation of femtoliter capillary bridges, we present a feasible experimental method to directly visualize how water bridges evaporate between a microsphere and a flat substrate in still air using transmission X-ray microscopy. Precise measurements of evaporation rates for water bridges show that lower water pressure than surrounding pressure can significantly decrease evaporation through the suppression of vapor diffusion. This finding provides insight into the evaporation of ultrasmall capillary bridges. PMID:26928329

Effect of Water Cut on Pressure Drop of Oil (D130) -Water Flow in 4″Horizontal Pipe

Science.gov (United States)

Basha, Mehaboob; Shaahid, S. M.; Al-Hems, Luai M.

2018-03-01

The oil-water flow in pipes is a challenging subject that is rich in physics and practical applications. It is often encountered in many oil and chemical industries. The pressure gradient of two phase flow is still subject of immense research. The present study reports pressure measurements of oil (D130)-water flow in a horizontal 4″ diameter stainless steel pipe at different flow conditions. Experiments were carried out for different water cuts (WC); 0-100%. Inlet oil-water flow rates were varied from 4000 to 8000 barrels-per-day in steps of 2000. It has been found that the frictional pressure drop decreases for WC = 0 - 40 %. With further increase in WC, friction pressure drop increases, this could be due to phase inversion.

Commercial spent nuclear fuel shipments in the United States, 1964--1987

International Nuclear Information System (INIS)

1990-12-01

This report provides an overview of US commercial light-water reactor spent-fuel shipments that have occurred from January, 1964 through December, 1987. A summary analysis was performed on these historical shipments, showing the amount of fuel that has been shipped to research facilities, reprocessing plants, away-from-reactor (AFR) storage sites, and other reactors. Also presented in this report is a listing of potential spent-fuel shipments to and/or from commercial nuclear plants. Table 1 provides the detailed listing of historical spent-fuel shipments. Table 2 is a summary of these shipments grouped by destination. Section IV discusses utility plans for future spent-fuel shipments. 2 tabs

Low cost sonoluminescence experiment in pressurized water

Science.gov (United States)

Bernal, L.; Insabella, M.; Bilbao, L.

2012-06-01

We present a low cost design for demostration and mesurements of light emmision from a sonoluminescence experiment. Using presurized water introduced in an acrylic cylinder and one piezoelectric from an ultrasonic cleaner, we are able to generate cavitacion zones with emission of light. The use of argon to pressurize the water improves the emission an the light can be seen at naked eye in a softlit ambient.

Validating criticality calculations for spent fuel with 252Cf-source-driven noise measurements

International Nuclear Information System (INIS)

Mihalczo, J.T.; Krass, A.W.; Valentine, T.E.

1992-01-01

The 252 Cf-Source-driven noise analysis method can be used for measuring the subcritical neutron multiplication factor k of arrays of spent light water reactor (LWR) fuel. This type of measurement provides a parameter that is directly related to the criticality state of arrays of LWR fuel. Measurements of this parameter can verify the criticality safety margins of spent LWR fuel configurations and thus could be a means of obtaining the information to justify burnup credit for spent LWR transportation/storage casks. The practicality of a measurement depends on the ability to install the hardware required to perform the measurement. Source chambers containing the 252 Cf at the required source intensity for this application have been constructed and have operated successfully for ∼10 years and can be fabricated to fit into control rod guide tubes of PWR fuel elements. Fission counters especially developed for spent-fuel measurements are available that would allow measurements of a special 3 x 3 spent fuel array and a typical burnup credit rail cask with spent fuel in unborated water. Adding a moderator around these fission counters would allow measurements with the typical burnup credit rail cask with borated water and the special 3 x 3 array with borated water. The recent work of Ficaro on modifying the KENO Va code to calculate by the Monte Carlo method the time sequences of pulses at two detectors near a fissile assembly from the fission chain multiplication process, initiated by a 252 Cf source in the assembly allows a direct computer calculation of the noise analysis data from this measurement method

Reuse of Hydrotreating Spent Catalyst

International Nuclear Information System (INIS)

Habib, A.M.; Menoufy, M.F.; Amhed, S.H.

2004-01-01

All hydro treating catalysts used in petroleum refining processes gradually lose activity through coking, poisoning by metal, sulfur or halides or lose surface area from sintering at high process temperatures. Waste hydrotreating catalyst, which have been used in re-refining of waste lube oil at Alexandria Petroleum Company (after 5 years lifetime) compared with the same fresh catalyst were used in the present work. Studies are conducted on partial extraction of the active metals of spent catalyst (Mo and Ni) using three leaching solvents,4% oxidized oxalic acid, 10% aqueous sodium hydroxide and 10% citric acid. The leaching experiments are conducting on the de coked extrude [un crushed] spent catalyst samples. These steps are carried out in order to rejuvenate the spent catalyst to be reused in other reactions. The results indicated that 4% oxidized oxalic acid leaching solution gave total metal removal 45.6 for de coked catalyst samples while NaOH gave 35% and citric acid gave 31.9 % The oxidized leaching agent was the most efficient leaching solvent to facilitate the metal removal, and the rejuvenated catalyst was characterized by the unchanged crystalline phase The rejuvenated catalyst was applied for hydrodesulfurization (HDS) of vacuum gas oil as a feedstock, under different hydrogen pressure 20-80 bar in order to compare its HDS activity

Complex cooling water systems optimization with pressure drop consideration

CSIR Research Space (South Africa)

Gololo, KV

2012-12-01

Full Text Available Pressure drop consideration has shown to be an essential requirement for the synthesis of a cooling water network where reuse/recycle philosophy is employed. This is due to an increased network pressure drop associated with additional reuse...

Containing method for spent fuel and spent fuel containing vessel

International Nuclear Information System (INIS)

Maekawa, Hiromichi; Hanada, Yoshine.

1996-01-01

Upon containing spent fuels, a metal vessel main body and a support spacer having fuel containing holes are provided. The support spacer is disposed in the inside of the metal vessel main body, and spent fuel assemblies are loaded in the fuel containing holes. Then, a lid is welded at the opening of the metal vessel main body to provide a sealing state. In this state, heat released from the spent fuel assemblies is transferred to the wall of the metal vessel main body via the support spacer. Since the support spacer has a greater heat conductivity than gases, heat of the spent fuel assemblies tends to be released to the outside, thereby capable of removing heat of the spent fuel assemblies effectively. In addition, since the surfaces of the spent fuel assemblies are in contact with the inner surface of the fuel containing holes of the support spacer, impact-resistance and earthquake-resistance are ensured, and radiation from the spent fuel assemblies is decayed by passing through the layer of the support spacer. (T.M.)

Light-water reactor pressure vessel surveillance standards

International Nuclear Information System (INIS)

Anon.

1981-01-01

The master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel steels throughout a pressure vessel's service life. Some of these are existing American Society for Testing and Materials (ASTM) standards, some are ASTM standards that have been modified, and some are newly proposed ASTM standards. The current (1) scope, (2) areas of application, (3) interrelationships, and (4) status and time table of development, improvement, validation, and calibration for a series of 16 ASTM standards are defined. The standard also includes a discussion of LWR pressure vessel surveillance - justification, requirements, and status of work

Evaluation of cover gas impurities and their effects on the dry storage of LWR [light-water reactor] spent fuel

International Nuclear Information System (INIS)

Knoll, R.W.; Gilbert, E.R.

1987-11-01

The purposes of this report are to (1) identify the sources of impurity gases in spent fuel storage casks; (2) identify the expected concentrations and types of reactive impurity gases from these sources over an operating lifetime of 40 years; and (3) determine whether these impurities could significantly degrade cladding or exposed fuel during this period. Four potential sources of impurity gases in the helium cover gas in operating casks were identified and evaluated. Several different bounding cases have been considered, where the reactive gas inventory is either assumed to be completely gettered by the cladding or where all oxygen is assumed to react completely with the exposed fuel. It is concluded that the reactive gas inventory will have no significant effect on the cladding unless all available oxygen reacts with the UO 2 fuel to produce U 3 O 8 at one or two cladding breaches. Based on Zircaloy oxidation data, the oxygen inventory in a fully loaded pressurized water reactor cask such as the Castor-V/21 will be gettered by the Zircaloy cladding in about 1 year if the peak cladding temperature within the task is ≥300 0 C. Only a negligible decrease in the thickness of the cladding would result. 24 refs., 4 tabs

Management of spent shea waste: An instrumental characterization and valorization in clay bricks construction.

Science.gov (United States)

Adazabra, A N; Viruthagiri, G; Shanmugam, N

2017-06-01

This work studies the reuse of spent shea waste as an economic construction material in improving fired clay bricks manufacture aside providing a novel approach to ecofriendly managing its excessive generated from the shea agroindustry. For this purpose, the influence of spent shea waste addition on the chemical, mineralogical, molecular bonding and technological properties (i.e. compressive strength and water absorption) of the fired clay bricks were extensively investigated. The results indicated that the chemical, mineralogical, phase transformations, molecular bonding and thermal behavior of the produced bricks were practically unaffected by the addition of spent shea waste. However, spent shea waste addition increased the compressive strengths and water absorptions of the brick products. Potential performance benefits of reusing spent shea waste was improved fluxing agents, energy-contribution reaction, excellent porosifying effect, reduced thermal conductivity and enhanced compressive strengths of the brick products. This research has therefore provided compelling evidence that could create newfound route for the synergistic ecofriendly reuse of spent shea waste to enhance clay brick construction aside being a potential mainstream disposal option. Copyright © 2017 Elsevier Ltd. All rights reserved.

High pressure, low pressure and hot water heating systems in hospitals. Hochdruck-, Niederdruck- und Warmwasserheizungsanlagen im Krankenhaus

Energy Technology Data Exchange (ETDEWEB)

Riedle, K [H. Riedle GmbH, Wiesbaden (Germany)

1994-07-01

In hospital nowadays the limitation of the use of steam boilers and their direct supply network to the possible minimum is aimed at when the heating system is exchanged or retrofitted. Independent of the fact whether high pressure or low pressure steam or hot water is used the optimum water treatment should be carried out with a minimum of chemical substances. Here hydroquinone, neutralizing amines, carbohydrazide, sodium sulphite and tannins can be used. The dimensioning of hot water heating circuits is shown with examples. (BWI)

Arrival condition of spent fuel after storage, handling, and transportation

International Nuclear Information System (INIS)

Bailey, W.J.; Pankaskie, P.J.; Langstaff, D.C.; Gilbert, E.R.; Rising, K.H.; Schreiber, R.E.

1982-11-01

This report presents the results of a study conducted to determine the probable arrival condition of spent light-water reactor (LWR) fuel after handling and interim storage in spent fuel storage pools and subsequent handling and accident-free transport operations under normal or slightly abnormal conditions. The objective of this study was to provide information on the expected condition of spent LWR fuel upon arrival at interim storage or fuel reprocessing facilities or at disposal facilities if the fuel is declared a waste. Results of a literature survey and data evaluation effort are discussed. Preliminary threshold limits for storing, handling, and transporting unconsolidated spent LWR fuel are presented. The difficulty in trying to anticipate the amount of corrosion products (crud) that may be on spent fuel in future shipments is also discussed, and potential areas for future work are listed. 95 references, 3 figures, 17 tables

On OH production in water containing atmospheric pressure plasmas

NARCIS (Netherlands)

Bruggeman, P.J.; Schram, D.C.

2010-01-01

In this paper radical production in atmospheric pressure water containing plasmas is discussed. As OH is often an important radical in these discharges the paper focuses on OH production. Besides nanosecond pulsed coronas and diffusive glow discharges, several other atmospheric pressure plasmas

The neutronics studies of a fusion fission hybrid reactor using pressure tube blankets

International Nuclear Information System (INIS)

Zheng Youqi; Zu Tiejun; Wu Hongchun; Cao Liangzhi; Yang Chao

2012-01-01

In this paper, a fusion fission hybrid reactor used for energy producing is proposed based on the situation of nuclear power in China. The pressurized light water is applied as the coolant. The fuel assemblies are loaded in the pressure tubes with a modular type structure. The neutronics analysis is performed to get the suitable design and prove the feasibility. The energy multiplication and tritium self-sustaining are evaluated. The neutron load is also cared. From different candidates, the PWR spent fuel is selected as the feed fuel. The results show that the hybrid reactor can meet the expected reactor core lifetime of 5 years with 1000 MWe power output. Two ways are discussed including burning the discharged PWR spent fuel and burning the reprocessed plutonium. The energy multiplication is big enough and the tritium can be self-sustaining for both of the two ways. The neutron wall load in the operating time is kept smaller than the one of ITER. The way to use the reprocessed plutonium brings low neutron wall load, but also brings additional difficulties in operating the hybrid reactor. The way to use the discharged spent fuel is proposed to be a better choice currently.

Failure Mode of the Water-filled Fractures under Hydraulic Pressure in Karst Tunnels

Directory of Open Access Journals (Sweden)

Dong Xin

2017-06-01

Full Text Available Water-filled fractures continue to grow after the excavation of karst tunnels, and the hydraulic pressure in these fractures changes along with such growth. This paper simplifies the fractures in the surrounding rock as flat ellipses and then identifies the critical hydraulic pressure values required for the occurrence of tensile-shear and compression-shear failures in water-filled fractures in the case of plane stress. The occurrence of tensile-shear fracture requires a larger critical hydraulic pressure than compression-shear failure in the same fracture. This paper examines the effects of fracture strike and lateral pressure coefficient on critical hydraulic pressure, and identifies compression-shear failure as the main failure mode of water-filled fractures. This paper also analyses the hydraulic pressure distribution in fractures with different extensions, and reveals that hydraulic pressure decreases along with the continuous growth of fractures and cannot completely fill a newly formed fracture with water. Fracture growth may be interrupted under the effect of hydraulic tensile shear.

Can the water content of highly compacted bentonite be increased by applying a high water pressure?

International Nuclear Information System (INIS)

Pusch, R.; Kasbohm, J.

2001-10-01

A great many laboratory investigations have shown that the water uptake in highly compacted MX-80 clay takes place by diffusion at low external pressure. It means that wetting of the clay buffer in the deposition holes of a KBS-3 repository is very slow if the water pressure is low and that complete water saturation can take several tens of years if the initial degree of water saturation of the buffer clay and the ability of the rock to give off water are low. It has therefore been asked whether injection of water can raise the degree of water saturation and if a high water pressure in the nearfield can have the same effect. The present report describes attempts to moisten highly compacted blocks of MX-80 clay with a dry density of 1510 kg/m 3 by injecting water under a pressure of 650 kPa through a perforated injection pipe for 3 and 20 minutes, respectively. The interpretation was made by determining the water content of a number of samples located at different distances from the pipe. An attempt to interpret the pattern of distribution of injected uranium acetate solution showed that the channels into which the solution went became closed in a few minutes and that dispersion in the homogenized clay gave low U-concentrations. The result was that the water content increased from about 9 to about 11-12 % within a distance of about 1 centimeter from the injection pipe and to slightly more than 9 % at a distance of about 4-5 cm almost independently of the injection time. Complete water saturation corresponds to a water content of about 30 % and the wetting effect was hence small from a practical point of view. By use of microstructural models it can be shown that injected water enters only the widest channels that remain after the compaction and that these channels are quickly closed by expansion of the hydrating surrounding clay. Part of the particles that are thereby released become transported by the flowing water and cause clogging of the channels, which is

Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods

International Nuclear Information System (INIS)

Johnson, G.L.

1991-11-01

Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. On such package would store tightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97 degrees C and whether the cladding of the stored spent fuel ever exceeds 350 degrees C. Limiting the borehole to temperatures of 97 degrees C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350 degrees C cladding limit minimizes the possibility of creep- related failure in the spent fuel rod cladding. For a series of packages stored in a 8 x 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97 degrees C for the full 10000-yr analysis period. For the 4.74-kW load, the peak cladding temperature rises to just below the 350 degrees C limit about 4 years after emplacement. If the packages are stored using the spacing specified in the Site Characterization Plan (15 ft x 126 ft), a maximum of 4.1 kW per container may be stored. If the 0.05-m-thick void between the container and the borehole wall is filled with loosely packed bentonite, the peak cladding temperature rises more than 40 degrees C above the allowed cladding limit. In all cases the dominant heat transfer mode between container components is thermal radiation

Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors

International Nuclear Information System (INIS)

Vito, D.J.

1980-12-01

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

Water cycle and its management for plant habitats at reduced pressures

Science.gov (United States)

Rygalov, Vadim Y.; Fowler, Philip A.; Wheeler, Raymond M.; Bucklin, Ray A.

2004-01-01

Experimental and mathematical models were developed for describing and testing temperature and humidity parameters for plant production in bioregenerative life support systems. A factor was included for analyzing systems operating at low (10-101.3 kPa) pressure to reduce gas leakage and structural mass (e.g., inflatable greenhouses for space application). The expected close relationship between temperature and relative humidity was observed, along with the importance of heat exchanger coil temperature and air circulation rate. The presence of plants in closed habitats results in increased water flux through the system. Changes in pressure affect gas diffusion rates and surface boundary layers, and change convective transfer capabilities and water evaporation rates. A consistent observation from studies with plants at reduced pressures is increased evapotranspiration rates, even at constant vapor pressure deficits. This suggests that plant water status is a critical factor for managing low-pressure production systems. The approach suggested should help space mission planners design artificial environments in closed habitats.

Design of virtual SCADA simulation system for pressurized water reactor

International Nuclear Information System (INIS)

Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

2016-01-01

The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor

Design of virtual SCADA simulation system for pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Wijaksono, Umar, E-mail: umar.wijaksono@student.upi.edu; Abdullah, Ade Gafar; Hakim, Dadang Lukman [Electrical Power System Research Group, Department of Electrical Engineering Education, Jl. Dr. Setiabudi No. 207 Bandung, Indonesia 40154 (Indonesia)

2016-02-08

The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

The Oxidation Rate of SiC in High Pressure Water Vapor Environments

Science.gov (United States)

Opila, Elizabeth J.; Robinson, R. Craig

1999-01-01

CVD SiC and sintered alpha-SiC samples were exposed at 1316 C in a high pressure burner rig at total pressures of 5.7, 15, and 25 atm for times up to 100h. Variations in sample emittance for the first nine hours of exposure were used to determine the thickness of the silica scale as a function of time. After accounting for volatility of silica in water vapor, the parabolic rate constants for Sic in water vapor pressures of 0.7, 1.8 and 3.1 atm were determined. The dependence of the parabolic rate constant on the water vapor pressure yielded a power law exponent of one. Silica growth on Sic is therefore limited by transport of molecular water vapor through the silica scale.

Comparison of air-charged and water-filled urodynamic pressure measurement catheters.

Science.gov (United States)

Cooper, M A; Fletter, P C; Zaszczurynski, P J; Damaser, M S

2011-03-01

Catheter systems are utilized to measure pressure for diagnosis of voiding dysfunction. In a clinical setting, patient movement and urodynamic pumps introduce hydrostatic and motion artifacts into measurements. Therefore, complete characterization of a catheter system includes its response to artifacts as well its frequency response. The objective of this study was to compare the response of two disposable clinical catheter systems: water-filled and air-charged, to controlled pressure signals to assess their similarities and differences in pressure transduction. We characterized frequency response using a transient step test, which exposed the catheters to a sudden change in pressure; and a sinusoidal frequency sweep test, which exposed the catheters to a sinusoidal pressure wave from 1 to 30 Hz. The response of the catheters to motion artifacts was tested using a vortex and the response to hydrostatic pressure changes was tested by moving the catheter tips to calibrated heights. Water-filled catheters acted as an underdamped system, resonating at 10.13 ± 1.03 Hz and attenuating signals at frequencies higher than 19 Hz. They demonstrated significant motion and hydrostatic artifacts. Air-charged catheters acted as an overdamped system and attenuated signals at frequencies higher than 3.02 ± 0.13 Hz. They demonstrated significantly less motion and hydrostatic artifacts than water-filled catheters. The transient step and frequency sweep tests gave comparable results. Air-charged and water-filled catheters respond to pressure changes in dramatically different ways. Knowledge of the characteristics of the pressure-measuring system is essential to finding the best match for a specific application. Copyright © 2011 Wiley-Liss, Inc.

Basal interstitial water pressure in laboratory debris flows over a rigid bed in an open channel

Directory of Open Access Journals (Sweden)

N. Hotta

2012-08-01

Full Text Available Measuring the interstitial water pressure of debris flows under various conditions gives essential information on the flow stress structure. This study measured the basal interstitial water pressure during debris flow routing experiments in a laboratory flume. Because a sensitive pressure gauge is required to measure the interstitial water pressure in shallow laboratory debris flows, a differential gas pressure gauge with an attached diaphragm was used. Although this system required calibration before and after each experiment, it showed a linear behavior and a sufficiently high temporal resolution for measuring the interstitial water pressure of debris flows. The values of the interstitial water pressure were low. However, an excess of pressure beyond the hydrostatic pressure was observed with increasing sediment particle size. The measured excess pressure corresponded to the theoretical excess interstitial water pressure, derived as a Reynolds stress in the interstitial water of boulder debris flows. Turbulence was thought to induce a strong shear in the interstitial space of sediment particles. The interstitial water pressure in boulder debris flows should be affected by the fine sediment concentration and the phase transition from laminar to turbulent debris flow; this should be the subject of future studies.

Spent-fuel pool thermal hydraulics: The evaporation question

International Nuclear Information System (INIS)

Yilmaz, T.P.; Lai, J.C.

1996-01-01

Many nuclear power plants are currently using dense fuel arrangements that increase the number of spent fuel elements stored in their spent-fuel pools (SFPs). The denser spent-fuel storage results in higher water temperatures, especially when certain event scenarios are analyzed. In some of these event scenarios, it is conservative to maximize the evaporation rate, while in other circumstances it is required to minimize the evaporation rates for conservatism. Evaporation is such a fundamental phenomenon that many branches of engineering developed various equations based on theory and experiments. The evaporation rates predicted by existing equations present a wide range of variation, especially at water temperatures >40 degrees C. Furthermore, a study on which equations provide the highest and lowest evaporation rates has not been done until now. This study explores the sensitivity of existing evaporation equations to various parameters and recommends the limiting evaporation equations for use in the solution of SFP thermal problems. Note that the results of this study may be applicable to a much wider range of applications from irrigation ponds, cooling lakes, and liquid-waste management to calculating adequate air exchange rate for swimming pools and health spas

Pyrochemical recovery of actinide elements from spent light water reactor fuel

International Nuclear Information System (INIS)

Johnson, G.K.; Pierce, R.D.; Poa, D.S.; McPheeters, C.C.

1994-01-01

Argonne National Laboratory is investigating salt transport and lithium pyrochemical processes for recovery of transuranic (TRU) elements from spent light water reactor fuel. The two processes are designed to recover the TRU elements in a form compatible with the Integral Fast Reactor (IFR) fuel cycle. The IFR is uniquely effective in consuming these long-lived TRU elements. The salt transport process uses calcium dissolved in Cu-35 wt % Mg in the presence of a CaCl 2 salt to reduce the oxide fuel. The reduced TRU elements are separated from uranium and most of the fission products by using a MgCl 2 transport salt. The lithium process, which does not employ a solvent metal, uses lithium in the presence of a LiCl salt as the reductant. After separation from the salt, the reduced metal is introduced into an electrorefiner, which separates the TRU elements from the uranium and fission products. In both processes, reductant and reduction salt are recovered by electrochemical decomposition of the oxide reaction product

Methodology for surge pressure evaluation in a water injection system

Energy Technology Data Exchange (ETDEWEB)

Meliande, Patricia; Nascimento, Elson A. [Universidade Federal Fluminense (UFF), Niteroi, RJ (Brazil). Dept. de Engenharia Civil; Mascarenhas, Flavio C.B. [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Lab. de Hidraulica Computacional; Dandoulakis, Joao P. [SHELL of Brazil, Rio de Janeiro, RJ (Brazil)

2009-07-01

Predicting transient effects, known as surge pressures, is of high importance for offshore industry. It involves detailed computer modeling that attempts to simulate the complex interaction between flow line and fluid in order to ensure efficient system integrity. Platform process operators normally raise concerns whether the water injection system is adequately designed or not to be protected against possible surge pressures during sudden valve closure. This report aims to evaluate the surge pressures in Bijupira and Salema water injection systems due to valve closure, through a computer model simulation. Comparisons among the results from empirical formulations are discussed and supplementary analysis for Salema system were performed in order to define the maximum volumetric flow rate for which the design pressure was able to withstand. Maximum surge pressure values of 287.76 bar and 318.58 bar, obtained in Salema and Bijupira respectively, using empirical formulations have surpassed the operating pressure design, while the computer model results have pointed the greatest surge pressure value of 282 bar in Salema system. (author)

Vegetative Propagule Pressure and Water Depth Affect Biomass and Evenness of Submerged Macrophyte Communities.

Science.gov (United States)

Li, Hong-Li; Wang, Yong-Yang; Zhang, Qian; Wang, Pu; Zhang, Ming-Xiang; Yu, Fei-Hai

2015-01-01

Vegetative propagule pressure may affect the establishment and structure of aquatic plant communities that are commonly dominated by plants capable of clonal growth. We experimentally constructed aquatic communities consisting of four submerged macrophytes (Hydrilla verticillata, Ceratophyllum demersum, Elodea nuttallii and Myriophyllum spicatum) with three levels of vegetative propagule pressure (4, 8 and 16 shoot fragments for communities in each pot) and two levels of water depth (30 cm and 70 cm). Increasing vegetative propagule pressure and decreasing water level significantly increased the growth of the submerged macrophyte communities, suggesting that propagule pressure and water depth should be considered when utilizing vegetative propagules to re-establish submerged macrophyte communities in degraded aquatic ecosystems. However, increasing vegetative propagule pressure and decreasing water level significantly decreased evenness of the submerged macrophyte communities because they markedly increased the dominance of H. verticillata and E. nuttallii, but had little impact on that of C. demersum and M. spicatum. Thus, effects of vegetative propagule pressure and water depth are species-specific and increasing vegetative propagule pressure under lower water level can facilitate the establishment success of submerged macrophyte communities.

Assessment of conditions of the spent nuclear fuel stored in the stainless steel channel-holders

International Nuclear Information System (INIS)

Pesic, M.; Sotic, O.; Cupac, S.; Maksin, T.; Dasic, N.

2003-01-01

The IAEA technical co-operation project 'Safe Removal of Spent Fuel of the Vinca RA Research Reactor' is carried out at the Vinca Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since January 2003. Present activities will provide up-to-date information on the conditions of the spent nuclear fuel, stored in the stainless steel channel-holders ('chekhols') and on the water quality in the storage basins. Water samples taken out from the chekhols and the basins are measured to determine their activity and chemical parameters. Until September 2003, about 1/3 of the chekhols containing spent fuel elements with initial enrichment of 2% and 80% of uranium were inspected. High activity of Cs-137 was found in several water samples taken out from chekhols. All water samples show very high electrical conductivity, while those taken from the basins show the presence of chlorides and aluminium ions, too. Information on established procedures and measuring results are given in this paper. The obtained results, so far, show that the spent nuclear fuel elements are leaking in about 10% of chekhols. (author)

Measurement of soluble nuclide dissolution rates from spent fuel

International Nuclear Information System (INIS)

Wilson, C.N.; Gray, W.J.

1990-01-01

Gaining a better understanding of the potential release behavior of water-soluble radionuclides is the focus of new laboratory spent fuel dissolution studies being planned in support of the Yucca Mountain Project. Previous studies have suggested that maximum release rates for actinide nuclides, which account for most of the long-term radioactivity in spent fuel, should be solubility-limited and should not depend on the characteristics or durability of the spent fuel waste form. Maximum actinide concentrations should be sufficiently low to meet the NRC (Nuclear Regulatory Commission) annual release limits. Potential release rates for soluble nuclides such as 99 Tc, 135 Cs, 14 C and 129 I, which account for about 1-2% of the activity in spent fuel at 1,000 years, are less certain and may depend on processes such as oxidation of the fuel in the repository air environment. Dissolution rates for several soluble nuclides have been measured from spent fuel specimens using static and semi-static methods. However, such tests do not provide a direct measurement of fuel matrix dissolution rates that may ultimately control soluble-nuclide release rates. Flow-through tests are being developed as a potential supplemental method for determining the matrix component of soluble-nuclide dissolution. Advantages and disadvantages of both semi-static and flow-through methods are discussed. Tests with fuel specimens representing a range of potential fuel states that may occur in the repository, including oxidized fuel, are proposed. Preliminary results from flow-through tests with unirradiated UO 2 suggesting that matrix dissolution rates are very sensitive to water composition are also presented

Processing of spent nuclear fuel from light water reactors

International Nuclear Information System (INIS)

Sraier, V.

1978-11-01

A comprehensive review is given of the reprocessing of spent nuclear fuel from LWR's (covering references up to No. 18 (1977) of INIS inclusively). Particular attention is devoted to waste processing, safety, and reprocessing plants. In the addendum, the present status is shown on the example of KEWA, the projected large German fuel reprocessing plant. (author)

Sloshing of water in torus pressure-suppression pool of boiling water reactors under earthquake ground motions

International Nuclear Information System (INIS)

Aslam, M.; Godden, W.G.; Scalise, D.T.

1978-08-01

This report presents an analytical and experimental investigation into the sloshing of water in torus tanks under horizontal earthquake ground motions. This study was motivated because of the use of torus tanks for pressure-suppression pools in Boiling Water Reactors. Such a pressure-suppression pool would typically have 80 ft and 140 ft inside and outside diameters, a 30 ft diameter section, and a water depth of 15 ft. A general finite element analysis was developed for all axisymmetric tanks and a computer program was written to obtain time-history plots of sloshing displacements of water and dynamic pressures. Tests were carried out on a 1/60th scale model under sinusoidal as well as simulated earthquake ground motions. Tests and analytical results regarding natural frequencies, surface water displacements, and dynamic pressures were compared and a good agreement was found within the range of displacements studied. The computer program gave satisfactory results within a maximum range of sloshing displacements in the full-size prototype of 30 in. which is greater than the value obtained under the full intensity of the El Centro earthquake (N-S component 1940). The range of linear behavior was studied experimentally by subjecting the torus model to increasing intensities of the El Centro earthquake

High converter pressurized water reactor with heavy water as a coolant

International Nuclear Information System (INIS)

Ronen, Y.; Reyev, D.

1983-01-01

There is an increasing interest in water breeder and high converter reactors. The increase in the conversion ratio of these reactors is obtained by hardening the neutron spectrum achieved by tightening the reactor's lattice. Another way of hardening the neutron spectrum is to replace the light water with heavy water. Two pressurized water reactor fuel cycles that use heavy water as a coolant are considered. The first fuel cycle is based on plutonium and depleted uranium, and the second cycle is based on plutonium and enriched uranium. The uranium ore and separative work unit (SWU) requirements are calculated as well as the fuel cycle cost. The savings in uranium ore are about40 and 60% and about40% in SWU for both fuel cycles considered

Spent fuel storage requirements. An update of DOE/RL-83-1

International Nuclear Information System (INIS)

1984-05-01

Spent fuel storage capacities at some commercial light water reactors (LWRs) are inadequate to handle projected spent fuel discharges. This report presents estimates of potential near-term requirements for additional LWR spent fuel storage capacity, based on information voluntarily supplied by utilities operating commercial nuclear power plants. These estimates provide information needed for planning the Department of Energy's (DOE) Federal Interim Storage (FIS) Program and the spent fuel research, development, and demonstration (RD and D) activities to be carried out under the DOE's Commercial Spent Fuel Management (CSFM) Program, in conjunction with the requirements of the Nuclear Waste Policy Act of 1982. This report is the latest in a series published by the DOE on LWR spent fuel storage requirements. Since the planning needs of the CSFM program focus on the near-term management of spent fuel inventories from commercial nuclear power reactors, the estimates in this report cover the ten-year period from the present through 1983. The report also assesses the possible impacts of using various concepts to reduce the requirements for additional storage capacity

Spent fuel storage capacities. An update of DOE/RL-84-1

International Nuclear Information System (INIS)

1985-10-01

Spent fuel storage capacities at some commercial light water reactors (LWRs) are inadequate to handle projected spent fuel discharges. This report presents estimates of potential near-term requirements for additional LWR spent fuel storage capacity, based on information supplied by utilities operating commercial nuclear power plants. These estimates provide information needed for planning the Department of Energy's (DOE) activities to be carried out under the DOE's Commercial Spent Fuel Management (CSFM) Program, in conjunction with the requirements of the Nuclear Waste Policy Act of 1982. The estimates in this report cover the period from the present through the year 2000. Although the DOE objective is to begin accepting spent fuel for final disposal in 1998, types of fuel and the receipt rates to be shipped are not yet known. Hence, this report makes no assumption regarding such fuel shipments. The resport also assesses the possible impacts of increased fuel exposure and spent fuel transhipment on the requirements for additional storage capacity

Reuse of spent granular activated carbon for organic micro-pollutant removal from treated wastewater.

Science.gov (United States)

Hu, Jingyi; Shang, Ran; Heijman, Bas; Rietveld, Luuk

2015-09-01

Spent granular activated carbons (sGACs) for drinking water treatments were reused via pulverizing as low-cost adsorbents for micro-pollutant adsorption from a secondary treated wastewater effluent. The changes of physicochemical characteristics of the spent carbons in relation to the fresh carbons were determined and were correlated to the molecular properties of the respective GAC influents (i.e. a surface water and a groundwater). Pore size distribution analysis showed that the carbon pore volume decreased over a wider size range due to preloading by surface water, which contains a broader molecular weight distribution of organic matter in contrast to the groundwater. However, there was still considerable capacity available on the pulverized sGACs for atrazine adsorption in demineralized water and secondary effluent, and this was particularly the case for the groundwater spent GAC. However, as compared to the fresh counterparts, the decreased surface area and the induced surface acidic groups on the pulverized sGACs contributed both to the lower uptake and the more impeded adsorption kinetic of atrazine in the demineralized water. Nonetheless, the pulverized sGACs, especially the one preloaded by surface water, was less susceptible to adsorption competition in the secondary effluent, due to its negatively charged surface which can repulse the accessibility of the co-present organic matter. This suggests the reusability of the drinking water spent GACs for micro-pollutant adsorption in the treated wastewater. Copyright © 2015 Elsevier Ltd. All rights reserved.

Effect of a cement buffer on spent fuel dissolution

International Nuclear Information System (INIS)

Mennecart, Thierry; Cachoir, Christelle; Lemmens, Karel; Gielen, Ben; Vercauter, Regina

2012-01-01

The Belgian agency for radioactive waste has selected the super-container design with an Ordinary Portland Cement (OPC) buffer as the reference design for geological disposal of High-Level Waste (HLW) and Spent Fuel (SF) in the Boom Clay formation. In the super-container design, the canisters of HLW or SF will be enclosed by a 30 mm thick carbon steel overpack and a 700 mm thick concrete buffer. The overpack will prevent contact with the (cementitious) pore water during the thermal phase. On the other hand, once the overpack will be locally perforated, the high pH of the incoming water may have an impact on the lifetime of the waste. Most published data and national programs are related to clayey backfill materials, and few studies are reported in alkaline media. Hence, a set of experiments was conducted to evaluate the behavior of spent fuel (UO 2 dissolution rate and UO 2 solubility) in such an environment. The objective was to estimate the spent fuel dissolution rate in super-container conditions for use in preliminary performance assessment calculations

Spent nuclear fuel discharges from US reactors 1993

International Nuclear Information System (INIS)

1995-02-01

The Energy Information Administration (EIA) of the U.S. Department of Energy (DOE) administers the Nuclear Fuel Data Survey, Form RW-859. This form is used to collect data on fuel assemblies irradiated at commercial nuclear reactors operating in the United States, and the current inventories and storage capacities of those reactors. These data are important to the design and operation of the equipment and facilities that DOE will use for the future acceptance, transportation, and disposal of spent fuels. The data collected and presented identifies trends in burnup, enrichment, and spent nuclear fuel discharged form commercial light-water reactor as of December 31, 1993. The document covers not only spent nuclear fuel discharges; but also site capacities and inventories; canisters and nonfuel components; and assembly type characteristics

Utilization of Spent Resources in Support of Eco-Economic Decoupling in Central Java

Directory of Open Access Journals (Sweden)

Nuril Fikri Aulia

2015-09-01

Full Text Available Implementation of the development is often cause adverse environmental impacts. Adverse effects are environmental degradation and decreasing availability of resources. To overcome this, it is necessary that the development can still continue, the environment is not damaged, and the availability of resources is maintained. One effort is through eco - economic decoupling activities with the use of spent resources. The aim of study to determine the potential of spent resources in Central Java, knows the problems in the utilization of spent resources in Central Java, and to determine the impact of the utilization of spent resources in Central Java by a qualitative descriptive method. The results show that in the study have the potential of eco-economic decoupling indicated by the availability of spent resources and had done utilization of spent resources. However, this potential has not been optimally developed, because there are still some problems in its utilization. Problems in the use of spent resources are the lack of knowledge about eco-economic decoupling and spent resources among stakeholder, there is no specific policy on eco - economic decoupling, the lack of Local Government 's role in the utilization of spent resource, and the lack of synergy programs and activities in supporting the utilization of spent resources. Utilization of spent resources have positive impact to reduce pressure on the environment and natural resources, create a new job, and increase incomes for society.

Steam generator for a pressurized-water coolant nuclear reactor

International Nuclear Information System (INIS)

Schroeder, H.J.; Berger, W.

1975-01-01

A description is given of a steam generator which has a vertical cylindrical housing having a steam output outlet, a horizontal tube sheet closing the lower end of this housing, and an inverted U-shaped tube bundle inside of the housing and having vertical inlet and outlet legs with their ends mounted in the tube sheet. Beneath the tube sheet there are inlet and outlet manifolds for the respective ends of the tube bundle so that pressurized-water coolant from a pressurized-water coolant nuclear reactor can be circulated through the tube bundle

Categorization of failed and damaged spent LWR [light-water reactor] fuel currently in storage

International Nuclear Information System (INIS)

Bailey, W.J.

1987-11-01

The results of a study that was jointly sponsored by the US Department of Energy and the Electric Power Research Institute are described in this report. The purpose of the study was to (1) estimate the number of failed fuel assemblies and damaged fuel assemblies (i.e., ones that have sustained mechanical or chemical damage but with fuel rod cladding that is not breached) in storage, (2) categorize those fuel assemblies, and (3) prepare this report as an authoritative, illustrated source of information on such fuel. Among the more than 45,975 spent light-water reactor fuel assemblies currently in storage in the United States, it appears that there are nearly 5000 failed or damaged fuel assemblies. 78 refs., 23 figs., 19 tabs

Time/motion observations and dose analysis of reactor loading, transportation, and dry unloading of an overweight truck spent fuel shipment

International Nuclear Information System (INIS)

Hostick, C.J.; Lavender, J.C.; Wakeman, B.H.

1992-04-01

This document presents observed activity durations and radiation dose analyses for an overweight truck shipment of pressurized water reactor (PWR) spent fuel from the Surry Power Station in Virginia to the Idaho National Engineering Laboratory. The shipment consisted of a TN-8L shipping cask carrying three 9-year-old PWR spent fuel assemblies. Handling times and dose analyses for at-reactor activities were completed by Virginia Electric and Power Company (Virginia Power) personnel. Observations of in-transit and unloading activities were made by Pacific Northwest Laboratory (PNL) personnel, who followed the shipment for approximately 2800 miles and observed cask unloading activities. In-transit dose estimates were calculated using dose rate maps provided by Virginia Power for a fully loaded TN-8L shipping cask. The dose analysis for the cask unloading operations is based on the observations of PNL personnel

Response of steam-water mixtures to pressure transients

International Nuclear Information System (INIS)

Hull, L.M.

1985-01-01

During the transition phase of a hypothetical core-disruptive accident in a liquid-metal fast breeder reactor, melting fuel-steel mixtures may begin to boil, resulting in a two-phase mixture of molten reactor fuel and steel vapor. Dispersal of this mixture by pressure transients may prevent recriticality of the fuel material. This paper describes the results of a series of experiments that investigated the response of two-phase mixtures to pressure transients. Simulant fluids (steam/water) were used in a transparent 10.2-cm-dia, 63.5-cm-long acrylic tube. The pressure transient was provided by releasing pressurized nitrogen from a supply tank. The data obtained are in the form of pressure-time records and high-speed movies. The varied parameters are initial void fraction (10% and 40%) and transient pressure magnitude (3.45 and 310 kPa)

Off-Stream Watering Systems and Partial Barriers as a Strategy to Maximize Cattle Production and Minimize Time Spent in the Riparian Area

Directory of Open Access Journals (Sweden)

Ashley A. Rawluk

2014-10-01

Full Text Available A study was conducted in 2009 at two locations in Manitoba (Killarney and Souris, Canada to determine the impact of off-stream waterers (OSW with or without natural barriers on (i amount of time cattle spent in the 10 m buffer created within the riparian area, referred to as the riparian polygon (RP, (ii watering location (OSW or stream, and (iii animal performance measured as weight gain. This study was divided into three 28-day periods over the grazing season. At each location, the pasture—which ranged from 21.0 ha to 39.2 ha in size—was divided into three treatments: no OSW nor barriers (1CONT, OSW with barriers along the stream bank to deter cattle from watering at the stream (2BARR, and OSW without barriers (3NOBARR. Cattle in 2BARR spent less time in the RP in Periods 1 (p = 0.0002, 2 (p = 0.1116, and 3 (p < 0.0001 at the Killarney site compared to cattle in 3NOBARR at the same site. Cattle in 2BARR at the Souris site spent more time in the RP in Period 1 (p < 0.0001 and less time in Period 2 (p = 0.0002 compared to cattle in 3NOBARR. Cattle did use the OSW, but not exclusively, as watering at the stream was still observed. The observed inconsistency in the effectiveness of the natural barriers on deterring cattle from the riparian area between periods and locations may be partly attributable to the environmental conditions present during this field trial as well as difference in pasture size and the ability of the established barriers to deter cattle from using the stream as a water source. Treatment had no significant effect (p > 0.05 on cow and calf weights averaged over the summer period. These results indicate that the presence of an OSW does not create significant differences in animal performance when used in extensive pasture scenarios such as those studied within the present study. Whereas the barriers did not consistently discourage watering at the stream, the results provide some indication of the efficacy of the OSW as well

Experimental Study on Peak Pressure of Shock Waves in Quasi-Shallow Water

Directory of Open Access Journals (Sweden)

Zhenxiong Wang

2015-01-01

Full Text Available Based on the similarity laws of the explosion, this research develops similarity requirements of the small-scale experiments of underwater explosions and establishes a regression model for peak pressure of underwater shock waves under experimental condition. Small-scale experiments are carried out with two types of media at the bottom of the water and for different water depths. The peak pressure of underwater shock waves at different measuring points is acquired. A formula consistent with the similarity law of explosions is obtained and an analysis of the regression precision of the formula confirms its accuracy. Significance experiment indicates that the influence of distance between measuring points and charge on peak pressure of underwater shock wave is the greatest and that of water depth is the least within the range of geometric parameters. An analysis of data from experiments with different media at the bottom of the water reveals an influence on the peak pressure, as the peak pressure of a shock wave in a body of water with a bottom soft mud and rocks is about 1.33 times that of the case where the bottom material is only soft mud.

Dissolution rates of aluminum-based spent fuels relevant to geological disposal

International Nuclear Information System (INIS)

Mickalonis, J.I.

2000-01-01

The Department of Energy is pursuing the option of direct disposal of a wide variety of spent nuclear fuels under its jurisdiction. Characterization of the various types of spent fuel is required prior to licensing by the Nuclear Regulatory Commission and acceptance of the fuel at a repository site. One category of required data is the expected rate of radionuclide and fissile release to the environment as a result of exposure to groundwater after closure of the repository. To provide this type of data for four different aluminum-based spent fuels, tests were conducted using a flow through method that allows the dissolution rate of the spent fuel matrix to be measured without interference by secondary precipitation reactions that would muddle interpretation of the results. Similar tests had been conducted earlier with light water reactor spent fuel, thereby allowing direct comparisons

Automated corrosion fatigue crack growth testing in pressurized water environments

International Nuclear Information System (INIS)

Ceschini, L.J.; Liaw, P.K.; Rudd, G.E.; Logsdon, W.A.

1984-01-01

This paper describes in detail a novel approach to construct a test facility for developing corrosion fatigue crack growth rate (FCGR) properties in aggressive environments. The environment studied is that of a pressurized water reactor (PWR) at 288 0 C (550 0 F) and 13.8 MPa (200 psig). To expedite data generation, each chamber was designed to accommodate two test specimens. A common water recirculation and pressurization system was employed to service two test chambers. Thus, four fatigue crack propagation rate tests could be conducted simultaneously in the pressurized water environment. The data analysis was automated to minimize the typically high labor costs associated with corrosion fatigue crack propagation testing. Verification FCGR tests conducted on an ASTM A469 rotor steel in a room temperature air environment as well as actual PWR environment FCGR tests performed on an ASTM A533 Grade B Class 2 pressure vessel steel demonstrated that the dual specimen test facility is an excellent system for developing the FCGR properties of materials in adverse environments

Correction of Pressure Drop in Steam and Water System in Performance Test of Boiler

Science.gov (United States)

Liu, Jinglong; Zhao, Xianqiao; Hou, Fanjun; Wu, Xiaowu; Wang, Feng; Hu, Zhihong; Yang, Xinsen

2018-01-01

Steam and water pressure drop is one of the most important characteristics in the boiler performance test. As the measuring points are not in the guaranteed position and the test condition fluctuation exsits, the pressure drop test of steam and water system has the deviation of measuring point position and the deviation of test running parameter. In order to get accurate pressure drop of steam and water system, the corresponding correction should be carried out. This paper introduces the correction method of steam and water pressure drop in boiler performance test.

Corrosion fatigue cracking behavior of Inconel 690 (TT) in secondary water of pressurized water reactors

International Nuclear Information System (INIS)

Xiao Jun; Chen Luyao; Qiu Shaoyu; Chen Yong; Lin Zhenxia; Fu Zhenghong

2015-01-01

Inconel 690 (TT) is one of the key materials for tubes of steam generators for pressurized water reactors, where it is susceptible to corrosion fatigue cracking. In this paper, the corrosion fatigue cracking behavior of Inconel 690 (TT) was investigated under small scale yielding conditions, in the simulated secondary water of pressurized water reactor. It was observed that the fatigue crack growth rate was accelerated by a maximum factor up to 3 in the simulated secondary water, comparing to that in room temperature air. In addition, it was found that the accelerating effect was influenced by out-of-plane cracking of corrosion fatigue cracks and also correlated with stress intensity factor range, maximum stress intensity factor and stress ratio. (authors)

Pressure Transient Model of Water-Hydraulic Pipelines with Cavitation

Directory of Open Access Journals (Sweden)

Dan Jiang

2018-03-01

Full Text Available Transient pressure investigation of water-hydraulic pipelines is a challenge in the fluid transmission field, since the flow continuity equation and momentum equation are partial differential, and the vaporous cavitation has high dynamics; the frictional force caused by fluid viscosity is especially uncertain. In this study, due to the different transient pressure dynamics in upstream and downstream pipelines, the finite difference method (FDM is adopted to handle pressure transients with and without cavitation, as well as steady friction and frequency-dependent unsteady friction. Different from the traditional method of characteristics (MOC, the FDM is advantageous in terms of the simple and convenient computation. Furthermore, the mechanism of cavitation growth and collapse are captured both upstream and downstream of the water-hydraulic pipeline, i.e., the cavitation start time, the end time, the duration, the maximum volume, and the corresponding time points. By referring to the experimental results of two previous works, the comparative simulation results of two computation methods are verified in experimental water-hydraulic pipelines, which indicates that the finite difference method shows better data consistency than the MOC.

Vegetative Propagule Pressure and Water Depth Affect Biomass and Evenness of Submerged Macrophyte Communities.

Directory of Open Access Journals (Sweden)

Hong-Li Li

Full Text Available Vegetative propagule pressure may affect the establishment and structure of aquatic plant communities that are commonly dominated by plants capable of clonal growth. We experimentally constructed aquatic communities consisting of four submerged macrophytes (Hydrilla verticillata, Ceratophyllum demersum, Elodea nuttallii and Myriophyllum spicatum with three levels of vegetative propagule pressure (4, 8 and 16 shoot fragments for communities in each pot and two levels of water depth (30 cm and 70 cm. Increasing vegetative propagule pressure and decreasing water level significantly increased the growth of the submerged macrophyte communities, suggesting that propagule pressure and water depth should be considered when utilizing vegetative propagules to re-establish submerged macrophyte communities in degraded aquatic ecosystems. However, increasing vegetative propagule pressure and decreasing water level significantly decreased evenness of the submerged macrophyte communities because they markedly increased the dominance of H. verticillata and E. nuttallii, but had little impact on that of C. demersum and M. spicatum. Thus, effects of vegetative propagule pressure and water depth are species-specific and increasing vegetative propagule pressure under lower water level can facilitate the establishment success of submerged macrophyte communities.

Comparison of the effects of crude and spent oils on Tilapia ...

African Journals Online (AJOL)

Toxic effects of bonnylight crude and spent oils on Oreochromis niloticus fingerlings were conducted under laboratory conditions at the University of Port Harcourt, Nigeria for two months. Five concentrations (5 ml/L, 10 ml/L, 20 m1/L, 50 m1/L and l00 ml/L) of water soluble fraction (WSF) of crude and spent oils were ...

Application-specific integrated circuit design for a typical pressurized water reactor pressure channel trip

International Nuclear Information System (INIS)

Battle, R.E.; Manges, W.W.; Emery, M.S.; Vendermolen, R.I.; Bhatt, S.

1994-01-01

This article discusses the use of application-specific integrated circuits (ASICs) in nuclear plant safety systems. ASICs have certain advantages over software-based systems because they can be simple enough to be thoroughly tested, and they can be tailored to replace existing equipment. An architecture to replace a pressurized water reactor pressure channel trip is presented. Methods of implementing digital algorithms are also discussed

Evaluation of economics of spent fuel storage techniques

International Nuclear Information System (INIS)

Yamaji, Kenji; Nagano, Koji

1988-01-01

Various spent fuel storage techniques are evaluated in terms of required costs. The unit storage cost for each spent fuel storage scenario is calculated based on the total cost required for the scenario including capital expenditure, operation cost, maintenance cost and transport cost. Intermediate storage may be performed in relatively small facilities in the plant or in independent large-scale facilities installed away from the plant. Dry casks or water pools are assumed to be used in in-plant storage facilities while vaults may also be employed in independent facilities. Evaluation is made for these different cases. In in-plant facilities, dry cask storage is found to be more economical in all cases than water pool storage, especially when large-sized casks are employed. In independent facilities, on the other hand, the use of vaults is the most desirable because the required capital expenditure is the lowest due to the effect of scale economics. Dry cask storage is less expensive than water pool storage also in independent facilities. The annual discount rate has relatively small influence on the unit cost for storage. An estimated unit cost for storage in independent storage facilities is shown separately for facilities with a capacity of 1,000 tons, 3,000 tons or 5,000 tons. The report also outlines the economics of spent fuel storage in overseas facilities (Finland, Sweden and U.S.A.). (Nogami, K.)

Investigation of temperature fluctuation phenomena in a stratified steam-water two-phase flow in a simulating pressurizer spray pipe of a pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Miyoshi, Koji, E-mail: miyoshi.koj@inss.co.jp; Takenaka, Nobuyuki; Ishida, Taisuke; Sugimoto, Katsumi

2017-05-15

Highlights: • Thermal hydraulics phenomena were discussed in a spray pipe of pressurizer. • Temperature fluctuation was investigated in a stratified steam-water two-phase. • Remarkable liquid temperature fluctuations were observed in the liquid layer. • The observed temperature fluctuations were caused by the internal gravity wave. • The temperature fluctuations decreased with increasing dissolved oxygen. - Abstract: Temperature fluctuation phenomena in a stratified steam-water two-phase flow in a horizontal rectangular duct, which simulate a pressurizer spray pipe of a pressurized water reactor, were studied experimentally. Vertical distributions of the temperature and the liquid velocity were measured with water of various dissolved oxygen concentrations. Large liquid temperature fluctuations were observed when the water was deaerated well and dissolved oxygen concentration was around 10 ppb. The large temperature fluctuations were not observed when the oxygen concentration was higher. It was shown that the observed temperature fluctuations were caused by the internal gravity wave since the Richardson numbers were larger than 0.25 and the temperature fluctuation frequencies were around the Brunt-Väisälä frequencies in the present experimental conditions. The temperature fluctuations decreased by the non-condensable gas since the non-condensable gas suppressed the condensation and the temperature difference in the liquid layer was small.

Spent-fuel storage: a private sector option

International Nuclear Information System (INIS)

Thomas, J.A.; Ross, S.R.

1983-01-01

The investigation was performed to delineate the legal and financial considerations for establishing private sector support for the planning and development of an independent spent-fuel storage facility (ISFSF). The preferred institutional structure was found to be one in which a not-for-profit corporation contracts with a limited partnership to handle the spent fuel. The limited partnership acquires the necessary land and constructs the ISFSF facility and then leases the facility to the not-for-profit corporation, which acquires spent-fuel rods from the utilities. The DOE must agree to purchase the spent-fuel rods at the expiration of term and warrant continued operation of the facility if policy changes at the federal level force the removal of the rods prior to completion of the contracted storage cycle. The DOE planning base estimate of spent-fuel storage requirements indicates a market potential adequate to support 10,000 MTU or more of spent-fuel storage prior to the time a government repository is available to accept spent fuel around the turn of the century. The estimated construction cost of a 5000-MTU water basin facility is $552 million. The total capital requirements to finance such a facility are estimated to be $695 million, based on an assumed capital structure of 70 percent debt and 30 percent equity. The estimated total levelized cost of storage, including operating costs, for the assumed 17-year life of the facility is $223 per kilogram of uranium. This is equivalent to a slightly less than one mill per kilowatt-hour increase in nuclear fuel costs at the nuclear power station that was the source of the spent fuel. In conclusion, within the context of the new Nuclear Waste Policy Act of 1982, the study points to both the need for and the advantages of private sector support for one or more ISFSFs and establishes a workable mechanism for the recovery of the costs of owning and operating such facilities. 3 figures, 4 tables

Pressure of drinking water network on the Meyrin site to be boosted

CERN Multimedia

2004-01-01

In the framework of the refurbishment of CERN's drinking water supply system, the final part of the network on the Meyrin site is to be connected to the pumping station operated by Services Industriels de Genève, bringing about a significant increase in the network pressure of up to 5 bar. This means that from January 2005 onwards, the water pressure in buildings will be increased from 2 - 4 bar to 7 - 9 bar. The TS Department will be checking and upgrading the drinking water supply equipment in toilets and washrooms. All users with devices connected to the water supply system are kindly requested to check that these are compatible with the new pressure levels. More information on the buildings affected, the new pressure levels and the dates on which the changes will come into effect can be found at: https://edms.cern.ch/document/525717/1 Should any equipment under your responsibility be incompatible with the future pressure levels, please contact the Technical Control Room on 72201.

Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

Energy Technology Data Exchange (ETDEWEB)

Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2011-07-01

This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

International Nuclear Information System (INIS)

Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R.

2011-01-01

This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

Studies and research concerning BNFP. Nuclear spent fuel transportation studies

International Nuclear Information System (INIS)

Anderson, R.T.; Maier, J.B.

1979-11-01

Currently, there are a number of institutional problems associated with the shipment of spent fuel assemblies from commercial nuclear power plants: new and conflicting regulations, embargoing of certain routes, imposition of transport safeguards, physical security in-transit, and a lack of definition of when and where the fuel will be moved. This report presents a summary of these types and kinds of problems. It represents the results of evaluations performed relative to fuel receipt at the Barnwell Nuclear Fuel Plant. Case studies were made which address existing reactor sites with near-term spent fuel transportation needs. Shipment by either highway, rail, water, or intermodal water-rail was considered. The report identifies the impact of new regulations and uncertainty caused by indeterminate regulatory policy and lack of action on spent fuel acceptance and storage. This stagnant situation has made it impossible for industry to determine realistic transportation scenarios for business planning and financial risk analysis. A current lack of private investment in nuclear transportation equipment is expected to further prolong the problems associated with nuclear spent fuel and waste disposition. These problems are expected to intensify in the 1980's and in certain cases will make continuing reactor plant operation difficult or impossible