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Sample records for pwr primary cooling

  1. Tribological study of hard coatings without cobalt intended to isolation components of PWR primary cooling system; Etude tribologique de revetements durs sans cobalt destines aux organes d`isolement du circuit primaire des REP

    Energy Technology Data Exchange (ETDEWEB)

    Cachon, L.

    1995-10-18

    The objective is to qualify coatings without cobalt to replace ``Stellites`` coatings in isolation valves of PWR primary cooling system, as Co is activated when passing in the reactor core and contaminated the cooling loop. Three families of coatings were tested: PVD thin films from 1 to 8 {mu}m monolayers of Cr/C{sub x} with x varying between 1.6 and 9.5 at% or multilayers of pure chromium and Cr/C{sub 1.6} at%, coatings with a thickness between 100 and 200 {mu}m of cermets NiCr{sub y} (y varying from 5 to 35 at%) matrix binding chromium or tungsten carbides, and thick coatings 2 mm thickness of cermets Nitronic 60 or Inconel 625 matrix binding 10, 20 or 30% titanium or niobium carbides. Stellite 6 (2 mm) is the reference coating for tribology. Coatings were qualified and selected by thermal shocks, corrosion and plane friction. The thin film and the thick families were disqualified by their destruction or by their high friction coefficient. Then coatings between 100 and 200 {mu}m were used in a valve mock-up working in PWR primary cooling system pressure and temperature conditions. Tests show that these coatings have better wear or tightness performances than stellite 6, except for a slightly higher friction coefficient. (A.B.).

  2. Activity transport models for PWR primary circuits; PWR-ydinvoimalaitoksen primaeaeripiirin aktiivisuuskulkeutumismallit

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    Tanner, V.; Rosenberg, R. [VTT Chemical Technology, Otaniemi (Finland)

    1995-03-01

    The corrosion products activated in the primary circuit form a major source of occupational radiation dose in the PWR reactors. Transport of corrosion activity is a complex process including chemistry, reactor physics, thermodynamics and hydrodynamics. All the mechanisms involved are not known and there is no comprehensive theory for the process, so experimental test loops and plant data are very important in research efforts. Several activity transport modelling attempts have been made to improve the water chemistry control and to minimise corrosion in PWR`s. In this research report some of these models are reviewed with special emphasis on models designed for Soviet VVER type reactors. (51 refs., 16 figs., 4 tabs.).

  3. Optimal design of passive containment cooling system for innovative PWR

    Directory of Open Access Journals (Sweden)

    Huiun Ha

    2017-08-01

    Full Text Available Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS of an innovative pressurized water reactor (PWR. A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT geometry, PCCS heat exchanger (PCCX location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed.

  4. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T. [Kansai Electric Power Company, Osaka (Japan); Shimizu, S.; Ogata, Y. [Mitsubishi Heavy Industries, Ltd., Kobe (Japan)

    1997-04-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years.

  5. Investigation of Burst Pressures in PWR Primary Pressure Boundary Components

    Directory of Open Access Journals (Sweden)

    Ihn Namgung

    2016-02-01

    Full Text Available In a reactor coolant system of a nuclear power plant (NPP, an overpressure protection system keeps pressure in the loop within 110% of design pressure. However if the system does not work properly, pressure in the loop could elevate hugely in a short time. It would be seriously disastrous if a weak point in the pressure boundary component bursts and releases radioactive material within the containment; and it may lead to a leak outside the containment. In this study, a gross deformation that leads to a burst of pressure boundary components was investigated. Major components in the primary pressure boundary that is structurally important were selected based on structural mechanics, then, they were used to study the burst pressure of components by finite element method (FEM analysis and by number of closed forms of theoretical relations. The burst pressure was also used as a metric of design optimization. It revealed which component was the weakest and which component had the highest margin to bursting failure. This information is valuable in severe accident progression prediction. The burst pressures of APR-1400, AP1000 and VVER-1000 reactor coolant systems were evaluated and compared to give relative margins of safety.

  6. Application of CATE 2.0 code on evaluating activated corrosion products in a PWR cooling loop

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Jingyu; Li, Lu; Chen, Yixue [North China Electric Power Univ., Beijing (China). School of Nuclear Science and Engineering

    2017-03-15

    In PWR plants, most Occupational Radiation Exposure (ORE) for personnel results from Activated Corrosion Products (ACPs) in the cooling loop. In order to evaluate the ACPs in the cooling loop, a three-region transport model is built up based on the theory of driving force from the concentration difference in CATE 2.0 code. In order to analyze the nuclide composition of ACPs, the EAF-2007 nuclear database is embedded in CATE 2.0. The case of MIT PCCL test loop is simulated to test the availability of CATE 2.0 on PWR ACPs evaluation, and the activity of Co-58 and Co-60 after operation for 42 days calculated by CATE 2.0 is consistent with that from the code CRUDSIM adopted by MIT. Then, the nuclide composition of ACPs is analyzed in detail respectively for operation of 42 days and 12 months using CATE 2.0. The results show that the short-lived nuclides contribute a majority of the activity in the regions of in-flux wall and coolant, while the long-lived nuclides contribute most of the activity in the region of out-flux wall.

  7. Surface Oxidation Phenomena of Ni-Based Alloy 600 in PWR Primary Water Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Yun Soo; Hwang, Seong Sik; Kim, Sung Woo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    There is, nevertheless, growing evidence in support for the internal oxidation model by Scot, in which grain boundary oxidation is responsible for embrittlement and cracking. Grain boundaries can act as an enhanced diffusion path for oxidation, and grain boundary oxidation can be regarded as a precursor for crack initiation. Oxidation of the grain boundary in almost all nickel-based alloys exposed to primary water is known to be detrimental for grin boundary cohesion. Panter et al. showed that the crack initiation time is strongly reduced when the specimens are pre-exposed in a simulated PWR environment in the absence of applied stress. The changes of the grain boundary structure and chemistry owing to oxygen penetration can increase the sensitivity to PWSCC under a load since grain boundary oxidization significantly weakens the grain boundary strength. Most of the important experimental results obtained are believed to correlate with the oxidation penetration into the material. A spinel structure was detected by XRD in the oxide layers. Several different types of oxide scales were found by SEM examination on the corroded surface of Alloy 600 after an immersion test in the primary water environments. Surface grain boundaries were oxidized by oxygen penetration into the matrix through grain boundaries. Grain boundary oxidization is thought to be the main reason for intergranular cracking in this alloy in a primary water environment of a PWR.

  8. MELCOR Modeling of Air-Cooled PWR Spent Fuel Assemblies in Water empty Fuel Pools

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L. E.; Lopez, C.

    2013-07-01

    The OECD Spent Fuel Project (SFP) investigated fuel degradation in case of a complete Loss-Of- Coolant-Accident in a PWR spent fuel pool. Analyses of the SFP PWR ignition tests have been conducted with the 1.86.YT.3084.SFP MELCOR version developed by SNL. The main emphasis has been placed on assessing the MELCOR predictive capability to get reasonable estimates of time-to-ignition and fire front propagation under two configurations: hot neighbor (i.e., adiabatic scenario) and cold neighbor (i.e., heat transfer to adjacent fuel assemblies). A detailed description of hypotheses and approximations adopted in the MELCOR model are provided in the paper. MELCOR results accuracy was notably different between both scenarios. The reasons are highlighted in the paper and based on the results understanding a set of remarks concerning scenarios modeling is given.

  9. Research on General Corrosion Property of 304L and 304NG Stainless Steels in Simulated PWR Primary Water

    Institute of Scientific and Technical Information of China (English)

    PENG; De-quan; HU; Shi-lin; ZHANG; Ping-zhu; WANG; Hui

    2012-01-01

    <正>The general corrosion behaviors of 304L and 304NG grade stainless steels in simulated pressurized water reactor (PWR) primary loop were studied using still autoclave, respectively, the corrosion test lasted for 1 680 hours. The corrosion oxide films were analyzed macroscopically and microscopically. The results are shown in Figs. 1, 2.

  10. PFM Analysis for Pre-Existing Cracks on Alloy 182 Weld in PWR Primary Water Environment using Monte Carlo Simulation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Phil; Bahn, Chi Bum [Pusan National University, Busan (Korea, Republic of)

    2015-10-15

    Probabilistic Fracture Mechanics (PFM) analysis was generally used to consider the scatter and uncertainty of parameters in complex phenomenon. Weld defects could be present in weld regions of Pressurized Water Reactors (PWRs), which cannot be considered by the typical fracture mechanics analysis. It is necessary to evaluate the effects of the pre-existing cracks in welds for the integrity of the welds. In this paper, PFM analysis for pre-existing cracks on Alloy 182 weld in PWR primary water environment was carried out using a Monte Carlo simulation. PFM analysis for pre-existing cracks on Alloy 182 weld in PWR primary water environment was carried out. It was shown that inspection decreases the gradient of the failure probability. And failure probability caused by the pre-existing cracks was stabilized after 15 years of operation time in this input condition.

  11. Replacement of Co-base alloy for radiation exposure reduction in the primary system of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Han, Jeong Ho; Nyo, Kye Ho; Lee, Deok Hyun; Lim, Deok Jae; Ahn, Jin Keun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Kim, Sun Jin [Hanyang Univ., Seoul (Korea, Republic of)

    1996-01-01

    Of numerous Co-free alloys developed to replace Co-base stellite used in valve hardfacing material, two iron-base alloys of Armacor M and Tristelle 5183 and one nickel-base alloy of Nucalloy 488 were selected as candidate Co-free alloys, and Stellite 6 was also selected as a standard hardfacing material. These four alloys were welded on 316SS substrate using TIG welding method. The first corrosion test loop of KAERI simulating the water chemistry and operation condition of the primary system of PWR was designed and fabricated. Corrosion behaviors of the above four kinds of alloys were evaluated using this test loop under the condition of 300 deg C, 1500 psi. Microstructures of weldment of these alloys were observed to identify both matrix and secondary phase in each weldment. Hardnesses of weld deposit layer including HAZ and substrate were measured using micro-Vickers hardness tester. The status on the technology of Co-base alloy replacement in valve components was reviewed with respect to the classification of valves to be replaced, the development of Co-free alloys, the application of Co-free alloys and its experiences in foreign NPPs, and the Co reduction program in domestic NPPs and industries. 18 tabs., 20 figs., 22 refs. (Author).

  12. LBB evaluation for a typical Japanese PWR primary loop by using the US NRC approved methods

    Energy Technology Data Exchange (ETDEWEB)

    Swamy, S.A.; Bhowmick, D.C.; Prager, D.E. [Westinghouse Nuclear Technology Division, Pittsburgh, PA (United States)

    1997-04-01

    The regulatory requirements for postulated pipe ruptures have changed significantly since the first nuclear plants were designed. The Leak-Before-Break (LBB) methodology is now accepted as a technically justifiable approach for eliminating postulation of double-ended guillotine breaks (DEGB) in high energy piping systems. The previous pipe rupture design requirements for nuclear power plant applications are responsible for all the numerous and massive pipe whip restraints and jet shields installed for each plant. This results in significant plant congestion, increased labor costs and radiation dosage for normal maintenance and inspection. Also the restraints increase the probability of interference between the piping and supporting structures during plant heatup, thereby potentially impacting overall plant reliability. The LBB approach to eliminate postulating ruptures in high energy piping systems is a significant improvement to former regulatory methodologies, and therefore, the LBB approach to design is gaining worldwide acceptance. However, the methods and criteria for LBB evaluation depend upon the policy of individual country and significant effort continues towards accomplishing uniformity on a global basis. In this paper the historical development of the U.S. LBB criteria will be traced and the results of an LBB evaluation for a typical Japanese PWR primary loop applying U.S. NRC approved methods will be presented. In addition, another approach using the Japanese LBB criteria will be shown and compared with the U.S. criteria. The comparison will be highlighted in this paper with detailed discussion.

  13. Precursor evolution and SCC initiation of cold-worked alloy 690 in simulated PWR primary water

    Energy Technology Data Exchange (ETDEWEB)

    Zhai, Ziqing; Kruska, Karen; Toloczko, Mychailo B.; Bruemmer, Stephen M.

    2017-03-27

    Stress corrosion crack initiation of two thermally-treated, cold-worked (CW) alloy 690 materials was investigated in 360oC simulated PWR primary water using constant load tensile (CLT) tests and blunt notch compact tension (BNCT) tests equipped with direct current potential drop (DCPD) for in-situ detection of cracking. SCC initiation was not detected by DCPD for the 21% and 31%CW CLT specimens loaded at their yield stress after ~9,220 h, however intergranular (IG) precursor damage and isolated surface cracks were observed on the specimens. The two 31%CW BNCT specimens loaded at moderate stress intensity after several cyclic loading ramps showed DCPD-indicated crack initiation after 10,400h exposure at constant stress intensity, which resulted from significant growth of IG cracks. The 21%CW BNCT specimens only exhibited isolated small IG surface cracks and showed no apparent DCPD change throughout the test. Interestingly, post-test cross-section examinations revealed many grain boundary (GB) nano-cavities in the bulk of all the CLT and BNCT specimens particularly for the 31%CW materials. Cavities were also found along GBs extending to the surface suggesting an important role in crack nucleation. This paper provides an overview of the evolution of GB cavities and will discuss their effects on crack initiation in CW alloy 690.

  14. SCC crack growth rate of cold-worked austenitic stainless steels in PWR primary water conditions

    Energy Technology Data Exchange (ETDEWEB)

    Guerre, C.; Raquet, O.; Herms, E. [Commissariat a l' Energie Atomique (CEA), DEN/DPC/SCCME/LECA, Gif-sur-Yvette Cedex (France); Marie, S. [Commissariat a l' Energie Atomique (CEA), DEN/DM2S/SEMT/LISN, Gif-sur-Yvette Cedex (France); Le Calvar, M. [Inst. for Radiological Protection and Nuclear Safety (IRSN), DSR/SAMS, Fontenay-aux-Roses Cedex (France)

    2007-07-01

    Stress corrosion cracking (SCC) of stainless steels (SS) is a significant cause of failure in the pressurized water reactors (PWR). Most of the reported case history failures of SS in PWR can be attributed to pollutants (chloride, sulphate) and / or locally oxygenated environments, even to sensitisation of the SS. However, some failures have been attributed to heavy cold work (CW) of SS. In laboratory tests, SCC initiation of cold-worked SS has been obtained using slow strain rate tests (SSRT) in nominal PWR environment. This paper describes constant load and cyclic crack growth rate (CGR) tests on cold-worked SS, on CT specimens. 304L and 316L have been tested with a CW up to 60 %. CW 316L is more prone to cracking than 304L. Over 30 % of CW, 316L is susceptible to crack propagation under constant load. CW is the main controlling parameter for cracking. (author))

  15. Power-Cooling-Mismatch Test Series Test PCM-7. Experiment operating specifications. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Sparks, D.T.; Smith, R.H.; Stanley, C.J.

    1979-02-01

    The experiment operating specifications for the Power-Cooling-Mismatch (PCM) Test PCM-7 to be conducted in the Power Burst Facility are described. The PCM Test Series was designed on the basis of a parametric evaluation of fuel behavior response with cladding temperature, rod internal pressure, time in film boiling, and test rod power being the variable parameters. The test matrix, defined in the PCM Experiment Requirements Document (ERD), encompasses a wide range of situations extending from pre-CHF (critical heat flux) PCMs to long duration operation in stable film boiling leading to rod failure.

  16. Study for highly functional resin (macroporous resin) superior in removing micro particles in PWR primary circuit: on-site test

    Energy Technology Data Exchange (ETDEWEB)

    Itou, A.; Kondo, K.; Kouzuma, Y., E-mail: ayumu_itou@kyuden.co.jp [Kyusyu Electric Power Co., Inc., Minami-ku, Fukuoka (Japan); Umehara, R.; Shimizu, Y., E-mail: Ruyji_Umehara@mhi.co.jp [Mitsubishi Heavy Industries, Ltd., Hyogo-ku, Kobe (Japan); Kogawa, N.; Nagamine, K., E-mail: nkogawa@ndc.hq.mhi.co.jp [Nuclear Development Corp., Tokaimura, Ibaraki (Japan)

    2010-07-01

    In Japanese PWR plants, efforts to remove particulate constituents containing radioactive cobalt which provides a source of radiation exposure, are needed. Performance evaluation study was conducted for macroporous resin which was said to possess excellent performance in removing particulate constituents and whose practical accomplishment at plants in USA was reported to be good. As one of the means for radiation exposure reduction in PWR, a study for application of crud removing resin to actual plant was executed by laboratory experiments using simulated crud (Fe{sub 3}O{sub 4} particle). In this study, following two mechanisms were demonstrated as the particle capturing mechanism of macroporous resin; physical trapping by fine pores on resin surface; electrical adsorption onto resin surface. In addition, in parallel to the study for application of macroporous resin to actual PWR plant, on-site study was planned to investigate the primary system water chemistry during various stages of actual plant operation and to research performance of particle capturing in detail. As the on-site study, column experiments, there water was let pass through the column, were planned for various operation stage (startup period, power operation period and shutdown period). A kind of conventional gel-type resin and three kinds of macroporous resin were examined for onsite tests. As to particulate capturing, basic knowledge regarding capturing efficiency and influence of water chemistry on capturing performance were ordered. Capturing performance of each resin tested became clear and was ordered by comparison. Effectiveness of macroporous resin with regard to crud removal in primary coolant was confirmed. (author)

  17. Characterization of Oxide Layer with Precipitates of HANA-6 Exposed in Simulated PWR Primary Water Environment

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    Jang, Hun; Lim, Jea Young; Lee, Sung Yong; Kim, Yoon Ho; Mok, Yong Kyoon [KEPCO NF, Daejeon (Korea, Republic of)

    2016-10-15

    The delayed oxidation behaviors of β-Nb ppts and their amorphization behaviors in HANA-6 and other Zr-base alloys have been frequently reported. On the other hand, although Zr(Nb,Fe)2 ppts could be formed in the HANA-6 alloy due to Fe impurities contained in Zrsponge, the oxidation behavior of Zr(Nb,Fe)2 ppts contained in HANA-6 alloy has not been fully understood. In this study, oxide characteristics of HANA-6 corroded in simulated PWR environment for 165 and 315 days were investigated. And, oxidation behaviors of Zr(Nb,Fe)2 ppts contained in HANA-6 alloy were investigated by TEM with EDS techniques. The superior corrosion property of HANA-6 has been confirmed through corrosion test in simulated PWR water for 387 days. By using TEM/EDS technique, the oxide characteristics with presence of β- Nb (or β-enriched), and ZrNbFe (possibly Zr(Nb,Fe){sub 2}) ppts have been characterized as follows. 1. Delayed oxidation behaviors of β-Nb and Zr(Nb,Fe){sub 2} ppts and their amorphization due to oxidation were observed from TEM/EDS analyses. 2. The oxide layers having crystallite and partially amorphous ppts were slightly increased with increasing corrosion test time from 165 days to 315 days. 3. In outer oxide layer, Fe in Zr(Nb,Fe){sub 2} ppt was depleted and dissolved to outer layer of ppt and bulk oxide layer.

  18. Characterization of interfacial reactions and oxide films on 316L stainless steel in various simulated PWR primary water environments

    Science.gov (United States)

    Chen, Junjie; Xiao, Qian; Lu, Zhanpeng; Ru, Xiangkun; Peng, Hao; Xiong, Qi; Li, Hongjuan

    2017-06-01

    The effect of water chemistry on the electrochemical and oxidizing behaviors of 316L SS was investigated in hydrogenated, deaerated and oxygenated PWR primary water at 310 °C. Water chemistry significantly influenced the electrochemical impedance spectroscopy parameters. The highest charge-transfer resistance and oxide-film resistance occurred in oxygenated water. The highest electric double-layer capacitance and constant phase element of the oxide film were in hydrogenated water. The oxide films formed in deaerated and hydrogenated environments were similar in composition but different in morphology. An oxide film with spinel outer particles and a compact and Cr-rich inner layer was formed in both hydrogenated and deaerated water. Larger and more loosely distributed outer oxide particles were formed in deaerated water. In oxygenated water, an oxide film with hematite outer particles and a porous and Ni-rich inner layer was formed. The reaction kinetics parameters obtained by electrochemical impedance spectroscopy measurements and oxidation film properties relating to the steady or quasi-steady state conditions in the time-period of measurements could provide fundamental information for understanding stress corrosion cracking processes and controlling parameters.

  19. Effect of water chemistry on environmentally assisted cracking of alloy 600 in simulated primary side PWR environments

    Energy Technology Data Exchange (ETDEWEB)

    Koenig, M. [Studsvik Nuvlear (Sweden); Lidar, P. [GSE Power Systems (Sweden); Engstroem, J. [Ringhals NPP (Sweden); Gott, K. [SKI Sweden (Sweden)

    2002-07-01

    Environmental aspects of crack growth due to intergranular stress corrosion cracking (IGSCC) of Alloy 600 in simulated primary side PWR environments have been studied. The purpose of the study was to quantify the effects of the water chemistry (Li, B and H{sub 2} concentrations, and the pH-value by adding KOH) on the crack growth rate, da/dt. 12.5 mm thick compact tension (CT) specimens were used for testing at a constant maximum stress intensity factor in the range of 26-32 MPa{open_square}m. The crack growth was continuously monitored using a direct current potential drop system. Intergranular crack growth due to IGSCC was dominant in the specimens, although there were also small fractions of transgranular cracking. Multivariate analysis was used on the results from the present work together with results from previous tests on the same material. Temperature and the stress intensity were also included as factors in the analysis. A partial least squares regression was developed and interaction effects between the factors were found to affect the crack growth rate. The Partial Least Square regression predicts the observed crack growth rates reasonably well. (authors)

  20. Effect of surface state on the oxidation behavior of welded 308L in simulated nominal primary water of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Ming, Hongliang; Zhang, Zhiming; Wang, Jiazhen; Zhu, Ruolin; Ding, Jie; Wang, Jianqiu, E-mail: wangjianqiu@imr.ac.cn; Han, En-Hou; Ke, Wei

    2015-05-15

    Highlights: • A duplex oxide film can be formed on the Welded 308L. • Surface state has no influence on the phase composition of the oxide film. • Surface state can affect the thickness of the oxide film. • Surface state can affect the morphology of the oxide film. - Abstract: The oxidation behavior of 308L weld metal (WM) with different surface state in the simulated nominal primary water of pressurized water reactor (PWR) was studied by scanning electron microscopy (SEM) equipped with energy dispersive X-ray spectroscopy (EDS), X-ray diffraction (XRD) analyzer and X-ray photoelectron spectroscopy (XPS). After 480 h immersion, a duplex oxide film composed of a Fe-rich outer layer (Fe{sub 3}O{sub 4}, Fe{sub 2}O{sub 3} and a small amount of NiFe{sub 2}O{sub 4}, Ni(OH){sub 2}, Cr(OH){sub 3} and (Ni, Fe)Cr{sub 2}O{sub 4}) and a Cr-rich inner layer (FeCr{sub 2}O{sub 4} and NiCr{sub 2}O{sub 4}) can be formed on the 308L WM samples with different surface state. The surface state has no influence on the phase composition of the oxide films but obviously affects the thickness of the oxide films and the morphology of the oxides (number & size). With increasing the density of dislocations and subgrain boundaries in the cold-worked superficial layer, the thickness of the oxide film, the number and size of the oxides decrease.

  1. Determination of the {sup 129}I in primary coolant of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Ke Chon; Park, Yong Joon; Song, Kyu Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-02-15

    Among the radioactive wastes generated from the nuclear power plant, a radioactive nuclide such as {sup 129}I is classified as a difficult-to-measure (DTM) nuclide, owing to its low specific activity. Therefore, the establishment of an analytical procedure, including a chemical separation for {sup 129}I as a representative DTM, becomes essential. In this report, the adsorption and recovery rate were measured by adding {sup 125}I as a radio-isotopic tracer (t1/2 = 60.14 d) to the simulation sample, in order to measure the activity concentration of {sup 129}I in a pressurized-water reactor primary coolant. The optimum condition for the maximum recovery yield of iodine on the anion exchange resins (AG1 x2, 50-100 mesh, Clform) was found to be at pH 7. In this report, the effect of the boron content in a pressurized-water reactor primary coolant on the separation process of {sup 129}I was examined, as was the effect of {sup 3}H on the measurement of the activity of iodine. As a result, no influence of the boron content and of the simultaneous {sup 3}H presence was found with activity concentrations of {sup 3}H lower than 50 Bq/mL, and with a boron concentration of less than 2,000 {mu}g/mL.

  2. Stress corrosion crack initiation of alloy 600 in PWR primary water

    Energy Technology Data Exchange (ETDEWEB)

    Zhai, Ziqing; Toloczko, Mychailo B.; Olszta, Matthew J.; Bruemmer, Stephen M.

    2017-07-01

    Stress corrosion crack (SCC) initiation of three mill-annealed (MA) alloy 600 heats in simulated pressurized water reactor primary water has been investigated using constant load tests equipped with in-situ direct current potential drop (DCPD) measurement capabilities. SCC initiation times were greatly reduced by a small amount of cold work. Shallow intergranular (IG) attack and/or cracks were found on most high-energy grain boundaries intersecting the surface with only a small fraction evolving into larger cracks and IGSCC growth. Crack depth profiles were measured and related to DCPD-detected initiation response. Processes controlling the SCC initiation in MA alloy 600 are discussed. IN PRESS, CORRECTED PROOF, 05/02/2017 - mfl

  3. Evaluation of the fuel rod integrity in PWR reactors from the spectrometric analysis of the primary coolant; Avaliacao da integridade de varetas combustiveis em reatores PWR a partir da analise espectrometrica da agua do primario

    Energy Technology Data Exchange (ETDEWEB)

    Monteiro, Iara Arraes

    1999-02-15

    The main objective of this thesis is to provide a better comprehension of the phenomena involved in the transport of fission products, from the fuel rod to the coolant of a PWR reactor. To achieve this purpose, several steps were followed. Firstly, it was presented a description of the fuel elements and the main mechanisms of fuel rod failure, indicating the most important nuclides and their transport mechanisms. Secondly, taking both the kinetic and diffusion models for the transport of fission products as a basis, a simple analytical and semi-empirical model was developed. This model was also based on theoretical considerations and measurements of coolant's activity, according to internationally adopted methodologies. Several factors are considered in the modelling procedures: intrinsic factors to the reactor itself, factors which depend on the reactor's operational mode, isotope characteristic factors, and factors which depend on the type of rod failure. The model was applied for different reactor's operational parameters in the presence of failed rods. The main conclusions drawn from the analysis of the model's output are relative to the variation on the coolant's water activity with the fuel burnup, the linear operation power and the primary purification rate and to the different behaviour of iodine and noble gases. The model was saturated from a certain failure size and showed to be unable to distinguish between a single big fail and many small ones. (author)

  4. Preoperational test report, primary ventilation condenser cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Clifton, F.T.

    1997-10-29

    This represents the preoperational test report for the Primary Ventilation Condenser Cooling System, Project W-030. Project W-030 provides a ventilation upgrade for the four Aging Waste Facility tanks. The system uses a closed chilled water piping loop to provide offgas effluent cooling for tanks AY101, AY102, AZ1O1, AZ102; the offgas is cooled from a nominal 100 F to 40 F. Resulting condensation removes tritiated vapor from the exhaust stack stream. The piping system includes a package outdoor air-cooled water chiller with parallel redundant circulating pumps; the condenser coil is located inside a shielded ventilation equipment cell. The tests verify correct system operation and correct indications displayed by the central Monitor and Control System.

  5. The Effects of Hot Bending on the Low Cycle Fatigue Behaviors of 347 SS in PWR Primary Environment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ho-Sub; Hong, Jong-Dae; Lee, Junho; Jang, Changheui [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-10-15

    Fatigue damage could be significant for some locations, especially the welds and bends where stress concentration is typically high. As a possible solution, a large radius hot-bending method has been suggested to eliminate some weld joints and all tight bends. However, for the hot-bending process which involves a high temperature thermal cycle, there is a concern about changes in mechanical properties including low cycle fatigue behaviors. In APR1400, Type 347 SS have been used as surge line pipes. Therefore, to verify the applicability of hot-bending on 347 SS surge line pipes, an environmental fatigue test program was initiated. In this paper, the preliminary results of the on-going test program are introduced. Also, the low cycle fatigue behaviors of 347 SS are compared with those of other grade of stainless steels. The effects of hot bending on the low cycle fatigue behavior of 347 SS were quantitatively evaluated. The fatigue life was compared with the estimated values per NUREG 6909 rev. 1. There are no distinct differences between NUREG 6909 and LCF tests. According to fractography and cross section analysis in progress, basically, the reduction of LCF life of 347 SS in PWR water was caused by operation of HIC mechanism. The cyclic stress responses shows that there is no secondary hardening in 330 .deg.C air and PWR water.

  6. Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

    Energy Technology Data Exchange (ETDEWEB)

    Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety

    2017-06-01

    The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

  7. Materials Reliability Program: Environmental Fatigue Testing of Type 304L Stainless Steel U-Bends in Simulated PWR Primary Water (MRP-137)

    Energy Technology Data Exchange (ETDEWEB)

    R.Kilian

    2004-12-01

    Laboratory data generated in the past decade indicate a significant reduction in component fatigue life when reactor water environmental effects are experimentally simulated. However, these laboratory data have not been supported by nuclear power plant component operating experience. In recent comprehensive review of laboratory, component and structural test data performed through the EPRI Materials Reliability Program, flow rate was identified as a critical variable that was generally not considered in laboratory studies but applicable in plant operating environments. Available data for carbon/low-alloy steel piping components suggest that high flow is beneficial regarding the effects of a reactor water environment. Similar information is lacking for stainless steel piping materials. This report documents progress made to date in an extensive testing program underway to evaluate the effects of flow rate on the corrosion fatigue of 304L stainless steel under simulated PWR primary water environmental conditions.

  8. Three dimensional calculations of the primary coolant flow in a 900 MW PWR vessel. Numerical simulation of the accurate RCP start-up flow rate

    Energy Technology Data Exchange (ETDEWEB)

    Martin, A.; Alvarez, D.; Cases, F.; Stelletta, S. [Electricite de France (EDF), 78 - Chatou (France). Lab. National d`Hydraulique

    1997-06-01

    This report explains the last results about the mixing in the 900 MW PWR vessels. The accurate fluid flow transient, induced by the RCP starting-up, is represented. In a first time, we present the Thermalhydraulic Finite Element Code N3S used for the 3D numerical computations. After that, results obtained for one reactor operation case are given. This case is dealing with the transient mixing of a clear plug in the vessel when one primary pump starts-up. A comparison made between two injection modes; a steady state fluid flow conditions or the accurate RCP transient fluid flow conditions. The results giving the local minimum of concentration and the time response of the mean concentration at the core inlet are compared. The results show the real importance of the unsteadiness characteristics of the fluid flow transport of the clear water plug. (author) 12 refs.

  9. Effects of cold work and stress on oxidation and SCC behavior of stainless steels in PWR primary water environments

    Energy Technology Data Exchange (ETDEWEB)

    Shoji, T.; Sakaguchi, K.; Lu, Z. [Fracture and Reliability Research Institute, Tohoku University, Sendai City 980-8579 (Japan); Hirano, S.; Hasegawa, Y. [Kansai Electric Power Co (Japan); Kobayashi, T.; Fujimoto, K.; Nomura, Y. [Mitsubishi Heavy Industries (Japan)

    2011-07-01

    Intergranular stress corrosion cracking (SCC) samples taken from a weld HAZ of 316 stainless steel welded to a low alloy steel of steam generator nozzle with nickel base alloy 82 in Mihama Unit 2 PWR plant were analyzed by extensive metallographic observation, micro-Raman spectroscopy, TEM analysis of stainless steel material, oxide morphology, compositional profiles as well as their crystal structures. The crack growth history during the plant operation is discussed in connection to a residual stress distribution at HAZ and distribution of oxides on/in the cracks. Possible time dependence of crack growth rate with crack growth in components was proposed based upon the evidences observed about oxides. The importance of surface integrity assessment in SCC initiation and propagation is emphasized from a point of view of oxidation localization which can be promoted by strain (dislocation density), straining and stress, which play a crucial role in oxidation due to accelerated mass transfer in oxides as well as underlying metallic materials. Especially, preferential oxidation along slip bands suggests that oxygen diffusion in such a region with a high dislocation density is faster than the other region. This fact implies that grain boundary can also be a preferential path of oxidation as has been observed by TEM, TOFSIMS and 3D-APT. This localization of oxidation and acceleration is discussed based upon an analysis of profile development at a stressed oxide/metal interface. The effects of environmental parameters, temperature, loading mode, and rolling procedures on SCC of stainless steels in simulated PWR environments were investigated by laboratory tests. Strong interactions among grain boundary structure, environmental parameters and interfacial oxidation kinetics, and SCC behavior are observed

  10. Visualized Research on Primary Loop Simulation for PWR Nuclear Power Plant%压水堆核电厂一回路仿真可视化研究

    Institute of Scientific and Technical Information of China (English)

    肖瑶; 巫英伟; 苏光辉; 秋穗正

    2013-01-01

    In this study the main equipments and the primary loop of PWR nuclear power plant (NPP) were analyzed in detail.The model of point neutron dynamics,steam generator model with two-phase drift-flux governing equations,3-zone nonequilibrium pressurizer model and 4-quadrant main pump performance model were established.Based on the above models,a NPP simulation program was developed by using mixed programming with FORTRAN90 and Visual C++.The simulation program is of capability to achieve visualized simulation for the main equipments in primary loop and entire system of PWR nuclear power plant.It provides not only the visualized functions of real-time plotting,zooming,etc.,but also the output of numerical results with standard picture and/or text formatting files.Besides,the program was validated by comparing the calculation results of the program developed by authors and those of RELAP5/MOD3.0.%对压水堆核电厂一回路系统及主要设备进行了详细分析,建立了点堆中子动力学模型、两相漂移流蒸汽发生器模型、三区不平衡稳压器模型和主循环泵四象限特性模型,并以此为基础使用FORTRAN90语言和Visual C++语言通过混合编程的方法开发了核电厂仿真分析程序,实现了对压水堆核电厂一回路主要设备及全系统的可视化仿真计算.软件提供实时绘图、缩放等可视化功能,还提供了数据结果的标准图片格式和标准文本格式输出.通过将程序的计算结果与RELAP5/MOD3.0计算结果进行比较,对程序的可靠性进行了验证.

  11. The role of Hydrogen and Creep in Intergranular Stress Corrosion Cracking of Alloy 600 and Alloy 690 in PWR Primary Water Environments ? a Review

    Energy Technology Data Exchange (ETDEWEB)

    Rebak, R B; Hua, F H

    2004-07-12

    Intergranular attack (IGA) and intergranular stress corrosion cracking (IGSCC) of Alloy 600 in PWR steam generator environment has been extensively studied for over 30 years without rendering a clear understanding of the essential mechanisms. The lack of understanding of the IGSCC mechanism is due to a complex interaction of numerous variables such as microstructure, thermomechanical processing, strain rate, water chemistry and electrochemical potential. Hydrogen plays an important role in all these variables. The complexity, however, significantly hinders a clearer and more fundamental understanding of the mechanism of hydrogen in enhancing intergranular cracking via whatever mechanism. In this work, an attempt is made to review the role of hydrogen based on the current understanding of grain boundary structure and chemistry and intergranular fracture of nickel alloys, effect of hydrogen on electrochemical behavior of Alloy 600 and Alloy 690 (e.g. the passive film stability, polarization behavior and open-circuit potential) and effect of hydrogen on PWSCC behavior of Alloy 600 and Alloy 690. Mechanistic studies on the PWSCC are briefly reviewed. It is concluded that further studies on the role of hydrogen on intergranular cracking in both inert and primary side environments are needed. These studies should focus on the correlation of the results obtained at different laboratories by different methods on materials with different metallurgical and chemical parameters.

  12. Experimental Investigation on the Effects of Coolant Concentration on Sub-Cooled Boiling and Crud Deposition on Reactor Cladding at Prototypical PWR Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Schultis, J., Kenneth; Fenton, Donald, L.

    2006-10-20

    Increasing demand for energy necessitates nuclear power units to increase power limits. This implies significant changes in the design of the core of the nuclear power units, therefore providing better performance and safety in operations. A major hindrance to the increase of nuclear reactor performance especially in Pressurized Deionized water Reactors (PWR) is Axial Offset Anomaly (AOA)--the unexpected change in the core axial power distribution during operation from the predicted distribution. This problem is thought to be occur because of precipitation and deposition of lithiated compounds like boric acid (H{sub 2}BO{sub 3}) and lithium metaborate (LiBO{sub 2}) on the fuel rod cladding. Deposited boron absorbs neutrons thereby affecting the total power distribution inside the reactor. AOA is thought to occur when there is sufficient build-up of crud deposits on the cladding during subcooled nucleate boiling. Predicting AOA is difficult as there is very little information regarding the heat and mass transfer during subcooled nucleate boiling. An experimental investigation was conducted to study the heat transfer characteristics during subcooled nucleate boiling at prototypical PWR conditions. Pool boiling tests were conducted with varying concentrations of lithium metaborate (LiBO{sub 2}) and boric acid (H{sub 2}BO{sub 3}) solutions in deionized water. The experimental data collected includes the effect of coolant concentration, subcooling, system pressure and heat flux on pool the boiling heat transfer coefficient. The analysis of particulate deposits formed on the fuel cladding surface during subcooled nucleate boiling was also performed. The results indicate that the pool boiling heat transfer coefficient degrades in the presence of boric acid and lithium metaborate compared to pure deionized water due to lesser nucleation. The pool boiling heat transfer coefficients decreased by about 24% for 5000 ppm concentrated boric acid solution and by 27% for 5000 ppm

  13. Physics of hydride fueled PWR

    Science.gov (United States)

    Ganda, Francesco

    The first part of the work presents the neutronic results of a detailed and comprehensive study of the feasibility of using hydride fuel in pressurized water reactors (PWR). The primary hydride fuel examined is U-ZrH1.6 having 45w/o uranium: two acceptable design approaches were identified: (1) use of erbium as a burnable poison; (2) replacement of a fraction of the ZrH1.6 by thorium hydride along with addition of some IFBA. The replacement of 25 v/o of ZrH 1.6 by ThH2 along with use of IFBA was identified as the preferred design approach as it gives a slight cycle length gain whereas use of erbium burnable poison results in a cycle length penalty. The feasibility of a single recycling plutonium in PWR in the form of U-PuH2-ZrH1.6 has also been assessed. This fuel was found superior to MOX in terms of the TRU fractional transmutation---53% for U-PuH2-ZrH1.6 versus 29% for MOX---and proliferation resistance. A thorough investigation of physics characteristics of hydride fuels has been performed to understand the reasons of the trends in the reactivity coefficients. The second part of this work assessed the feasibility of multi-recycling plutonium in PWR using hydride fuel. It was found that the fertile-free hydride fuel PuH2-ZrH1.6, enables multi-recycling of Pu in PWR an unlimited number of times. This unique feature of hydride fuels is due to the incorporation of a significant fraction of the hydrogen moderator in the fuel, thereby mitigating the effect of spectrum hardening due to coolant voiding accidents. An equivalent oxide fuel PuO2-ZrO2 was investigated as well and found to enable up to 10 recycles. The feasibility of recycling Pu and all the TRU using hydride fuels were investigated as well. It was found that hydride fuels allow recycling of Pu+Np at least 6 times. If it was desired to recycle all the TRU in PWR using hydrides, the number of possible recycles is limited to 3; the limit is imposed by positive large void reactivity feedback.

  14. Tensile and Fatigue Testing and Material Hardening Model Development for 508 LAS Base Metal and 316 SS Similar Metal Weld under In-air and PWR Primary Loop Water Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish [Argonne National Lab. (ANL), Argonne, IL (United States); Soppet, William [Argonne National Lab. (ANL), Argonne, IL (United States); Majumdar, Saurin [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, Ken [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-09-01

    This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in September 2015 under the work package for environmentally assisted fatigue under DOE’s Light Water Reactor Sustainability program. In an April 2015 report we presented a baseline mechanistic finite element model of a two-loop pressurized water reactor (PWR) for systemlevel heat transfer analysis and subsequent thermal-mechanical stress analysis and fatigue life estimation under reactor thermal-mechanical cycles. In the present report, we provide tensile and fatigue test data for 508 low-alloy steel (LAS) base metal, 508 LAS heat-affected zone metal in 508 LAS–316 stainless steel (SS) dissimilar metal welds, and 316 SS-316 SS similar metal welds. The test was conducted under different conditions such as in air at room temperature, in air at 300 oC, and under PWR primary loop water conditions. Data are provided on materials properties related to time-independent tensile tests and time-dependent cyclic tests, such as elastic modulus, elastic and offset strain yield limit stress, and linear and nonlinear kinematic hardening model parameters. The overall objective of this report is to provide guidance to estimate tensile/fatigue hardening parameters from test data. Also, the material models and parameters reported here can directly be used in commercially available finite element codes for fatigue and ratcheting evaluation of reactor components under in-air and PWR water conditions.

  15. Enhanced Control of PWR Primary Coolant Water Chemistry Using Selective Separation Systems for Recovery and Recycle of Enriched Boric Acid

    Energy Technology Data Exchange (ETDEWEB)

    Ken Czerwinski; Charels Yeamans; Don Olander; Kenneth Raymond; Norman Schroeder; Thomas Robison; Bryan Carlson; Barbara Smit; Pat Robinson

    2006-02-28

    The objective of this project is to develop systems that will allow for increased nuclear energy production through the use of enriched fuels. The developed systems will allow for the efficient and selective recover of selected isotopes that are additives to power water reactors' primary coolant chemistry for suppression of corrosion attack on reactor materials.

  16. Enhanced Control of PWR Primary Coolant Water Chemistry Using Selective Separation Systems for Recovery and Recycle of Enriched Boric Acid

    Energy Technology Data Exchange (ETDEWEB)

    Ken Czerwinski; Charels Yeamans; Don Olander; Kenneth Raymond; Norman Schroeder; Thomas Robison; Bryan Carlson; Barbara Smit; Pat Robinson

    2006-02-28

    The objective of this project is to develop systems that will allow for increased nuclear energy production through the use of enriched fuels. The developed systems will allow for the efficient and selective recover of selected isotopes that are additives to power water reactors' primary coolant chemistry for suppression of corrosion attack on reactor materials.

  17. 压水堆主管道上充管嘴弹塑性应力分析%Elastoplastic Stress Analysis for Charging Nozzle of PWR Primary Piping

    Institute of Scientific and Technical Information of China (English)

    艾红雷; 郑斌; 卢喜丰; 王新军

    2016-01-01

    压水堆主管道上充管嘴在核电厂运行期间需经受严厉的冷热流体交互流动产生的循环热载荷,将对上充管嘴的结构完整性产生重要的影响.上充管嘴弹性应力分析证明了结构不会出现弹性失稳、塑性失稳以及疲劳破坏等现象,但部分分析截面一次加二次应力强度范围超过了规范限值.针对弹性分析部分结果不满足规范限值的情况,对上充管嘴进行了循环弹塑性分析,结果表明,上充管嘴结构在循环载荷作用下出现了明显的塑性安定现象,并且经历所分析的循环载荷后,其结构的累积应变不会对结构抗塑性失稳能力和抗疲劳破坏能力产生显著的影响.上充管嘴抗快断分析表明,其结构具备良好的抗快断性能.%The severe cyclical thermal loadings which generated by the interaction flow of hot and cold flu-id are subjected to the charging nozzle of PWR primary piping during the operation of nuclear power plants.The loadings have significant impact on the structural integrity.The types of damage( elastic insta-bility,plastic instability,fatigue and so on) do not occur for the nozzle is proved by elastic analysis.For some analysis sections,the range of primary plus secondary stresses cannot be validated and the cyclical elastoplastic analysis is used.The results show that the behavior of plastic accommodation occurred under the cyclical loadings and the cumulative strain induced by repeated loads do not diminish the capacity to prevent the damage of plastic instability and fatigue.The results of resistance to fast fracture analysis show that the nozzle has a good ability of resistance to fast fracture.

  18. Stress corrosion cracking of Ni-based alloys in PWR primary water. Component surface control; Corrosion sous contrainte des alliages a base nickel en milieu primaire des reacteurs a eaux pressurisee. Maitrise de la surface des composants

    Energy Technology Data Exchange (ETDEWEB)

    Foucault, M. [AREVA, Centre Technique Framatome ANP, Dept. Corrosion Chimie, 71 - Le Creusot (France)

    2004-06-01

    In the PWR plant primary circuit, FRAMATOME-ANP uses several nickel-base alloys or austenitic stainless steels for the manufacture of safety components. The experience feedback of the last twenty years allows us to point out the major role played by the surface state of the components in their life duration. In this paper, we present two examples of problems encountered and solved by a surface study and the definition and implementation of a process for the surface control of the repair components. Then, we propose some ideas about the present needs in terms of analysis methods to improve the surface knowledge and the control of the manufactured components. (author)

  19. Effects of dissolved hydrogen on general corrosion behavior and oxide films of alloy 690TT in PWR primary water

    Science.gov (United States)

    Jeon, Soon-Hyeok; Lee, Eun-Hee; Hur, Do Haeng

    2017-03-01

    The effect of dissolved hydrogen (DH) on the general corrosion behavior and oxide films of Alloy 690TT is investigated in simulated primary water at 330 °C. With increasing DH, the structure of oxide film significantly changed and the corrosion rate decreased. At DH = 5 cm3/kg H2O, the oxide layer was thick, and consisted of outer Ni oxide layer and inner Cr2O3 layer. Under the conditions of DH = 35 and 100 cm3/kg H2O, the oxide films grew thinner and composed of outer polyhedral spinel oxide particles such as NiCr2O4 or NiCrFeO4 and an intermediate metallic Ni-rich layer, with inner Cr2O3 layer. The general corrosion rate significantly decreased by about 72% as DH concentration increased from 5 to 35 cm3/kg H2O. In the range of 35-65 cm3/kg H2O, the corrosion rate slightly decreased with increasing DH concentration. However, no further changes were observed in the range of 65-100 cm3/kg H2O.

  20. Standard PWR for Italy

    Energy Technology Data Exchange (ETDEWEB)

    Negroni, A.; Velona, F. (Ente Nazionale per l' Energia Elettrica, Rome (Italy))

    1983-03-01

    A description is given of the general design for the standard PWR which will be used in the seven to eight nuclear power stations provided for in the Italian national energy plan. Special features to meet Italian conditions include double containment and a common foundation mat for the reactor, auxiliary and fuel buildings.

  1. PWR and BWR spent fuel assembly gamma spectra measurements

    Science.gov (United States)

    Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

    2016-10-01

    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 137Cs, 154Eu, and 134Cs 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.

  2. CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

    OpenAIRE

    Pengcheng Zhao; Kangli Shi; Shuzhou Li; Jingchao Feng; Hongli Chen

    2016-01-01

    Small modular reactor (SMR) has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR) is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100) is being developed by University of Science and Technology of China (USTC). In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kineti...

  3. The effect of primary copper slag cooling rate on the copper valorization in the flotation process

    Directory of Open Access Journals (Sweden)

    Aleksandar Mihajlović

    2015-06-01

    Full Text Available Technological procedure of slow cooling slag from primary copper production is applied in the purpose of copper recovery in the level of 98.5% to blister. This technological procedure is divided into two phases, first slow cooling of slag on the air for 24 hours, and then accelerated cooling with water for 48 hours. Within the research following methods were used: calculation of nonstationary slag cooling, verification of the calculation using computer simulation of slag cooling in the software package COMSOL Multiphysics and experimental verification of simulation results. After testing of the experimentally gained samples of slowly cooled slag it was found that this technological procedure gives the best results in promoting growth or coagulation of dispersed particles of copper sulfide and copper in the slag, thereby increasing the utilization of the flotation process with a decrease of copper losses through very fine particles.

  4. Influence of localized deformation on A-286 austenitic stainless steel stress corrosion cracking in PWR primary water; Influence de la localisation de la deformation sur la corrosion sous contrainte de l'acier inoxydable austenitique A-286 en milieu primaire des REP

    Energy Technology Data Exchange (ETDEWEB)

    Savoie, M

    2007-01-15

    Irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels is known to be a critical issue for structural components of nuclear reactor cores. The deformation of irradiated austenitic stainless steels is extremely heterogeneous and localized in deformation bands that may play a significant role in IASCC. In this study, an original approach is proposed to determine the influence of localized deformation on austenitic stainless steels SCC in simulated PWR primary water. The approach consists in (i) performing low cycle fatigue tests on austenitic stainless steel A-286 strengthened by {gamma}' precipitates Ni{sub 3}(Ti,Al) in order to shear and dissolve the precipitates in intense slip bands, leading to a localization of the deformation within and in (ii) assessing the influence of these {gamma}'-free localized deformation bands on A-286 SCC by means of comparative CERT tests performed on specimens with similar yield strength, containing or not {gamma}'-free localized deformation bands. Results show that strain localization significantly promotes A-286 SCC in simulated PWR primary water at 320 and 360 C. Moreover, A-286 is a precipitation-hardening austenitic stainless steel used for applications in light water reactors. The second objective of this work is to gain insights into the influence of heat treatment and metallurgical structure on A-286 SCC susceptibility in PWR primary water. The results obtained demonstrate a strong correlation between yield strength and SCC susceptibility of A-286 in PWR primary water at 320 and 360 C. (author)

  5. CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

    Directory of Open Access Journals (Sweden)

    Pengcheng Zhao

    2016-01-01

    Full Text Available Small modular reactor (SMR has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100 is being developed by University of Science and Technology of China (USTC. In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS transient simulation at beginning of the reactor cycle (BOC has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.

  6. PWR decontamination feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    Silliman, P.L.

    1978-12-18

    The decontamination work which has been accomplished is reviewed and it is concluded that it is worthwhile to investigate further four methods for decontamination for future demonstration. These are: dilute chemical; single stage strong chemical; redox processes; and redox/chemical in combination. Laboratory work is recommended to define the agents and processes for demonstration and to determine the effect of the solvents on PWR materials. The feasibility of Indian Point 1 for decontamination demonstrations is discussed, and it is shown that the system components of Indian Point 1 are well suited for use in demonstrations.

  7. Leak before break application in French PWR plants under operation

    Energy Technology Data Exchange (ETDEWEB)

    Faidy, C. [EDF SEPTEN, Villeurbanne (France)

    1997-04-01

    Practical applications of the leak-before break concept are presently limited in French Pressurized Water Reactors (PWR) compared to Fast Breeder Reactors. Neithertheless, different fracture mechanic demonstrations have been done on different primary, auxiliary and secondary PWR piping systems based on similar requirements that the American NUREG 1061 specifications. The consequences of the success in different demonstrations are still in discussion to be included in the global safety assessment of the plants, such as the consequences on in-service inspections, leak detection systems, support optimization,.... A large research and development program, realized in different co-operative agreements, completes the general approach.

  8. Metallurgical and mechanical parameters controlling alloy 718 stress corrosion cracking resistance in PWR primary water; Facteurs metallurgiques et mecaniques controlant l'amorcage de defauts de corrosion sous contrainte dans l'alliage 718 en milieu primaire des reacteurs a eau sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Deleume, J

    2007-11-15

    Improving the performance and reliability of the fuel assemblies of the pressurized water reactors requires having a perfect knowledge of the operating margins of both the components and the materials. The choice of alloy 718 as reference material for this study is justified by the industrial will to identify the first order parameters controlling the excellent resistance of this alloy to Stress Corrosion Cracking (SCC). For this purpose, a specific slow strain rate (SSR) crack initiation test using tensile specimen with a V-shaped hump in the middle of the gauge length was developed and modeled. The selectivity of such SSR tests in simulated PWR primary water at 350 C was clearly established by characterizing the SCC resistance of nine alloy 718 thin strip heats. Regardless of their origin and in spite of a similar thermo-mechanical history, they did not exhibit the same susceptibility to SCC crack initiation. All the characterized alloy 718 heats develop oxide scale of similar nature for various exposure times to PWR primary medium in the temperature range [320 C - 360 C]. {delta} phase precipitation has no impact on alloy 718 SCC initiation behavior when exposed to PWR primary water, contrary to interstitial contents and the triggering of plastic instabilities (PLC phenomenon). (author)

  9. Evaluation of alternative descriptions of PWR cladding corrosion behavior

    Energy Technology Data Exchange (ETDEWEB)

    Quecedo, M.; Serna, J. J.; Weiner, R. A.; Kersting, P. J.

    1999-05-15

    A statistical procedure has been used to evaluate several alternative descriptions of pressurized water reactor (PWR) cladding corrosion behavior, using an extensive database of Improved (low tin) Zr-4 cladding corrosion measurements from fuel irradiated in commercial PWRs. The in-reactor corrosion enhancement factors considered in the model development are based on a comprehensive review of the current literature for PWR cladding corrosion phenomenology and models. In addition, because prediction of PWR cladding corrosion behavior is very sensitive to the values used for the oxide surface temperatures, several models for the forced convection and sub-cooled nucleate boiling (SNB) coolant heat transfer under PWR conditions have also been evaluated. This evaluation determined that the choice of the forced convection heat transfer has the greatest impact on the ability to fit the data. In addition, the SNB heat transfer model used must account for a continuous transition from forced convection conditions to fully developed SNB conditions. With these choices for the heat transfer models, the evaluation determined that the significant in-reactor corrosion enhancement factors are related to the formation of a hydride rim at the cladding outer diameter, the coolant lithium concentration, and the fast neutron fluence (author) (ml)

  10. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    Directory of Open Access Journals (Sweden)

    Virpi Kouhia

    2012-01-01

    Full Text Available This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  11. Modeling in fast dynamics of accidents in the primary circuit of PWR type reactors; Modelisation en dynamique rapide d'accidents dans le circuit primaire des reacteurs a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Robbe, M.F

    2003-12-01

    Two kinds of accidents, liable to occur in the primary circuit of a Pressurized Water Reactor and involving fast dynamic phenomena, are analyzed. The Loss Of Coolant Accident (LOCA) is the accident used to define the current PWR. It consists in a large-size break located in a pipe of the primary circuit. A blowdown wave propagates through the circuit. The pressure differences between the different zones of the reactor induce high stresses in the structures of the lower head and may degrade the reactor core. The primary circuit starts emptying from the break opening. Pressure decreases very quickly, involving a large steaming. Two thermal-hydraulic simulations of the blowdown phase of a LOCA are computed with the Europlexus code. The primary circuit is represented by a pipe-model including the hydraulic peculiarities of the circuit. The main differences between both computations concern the kind of reactor, the break location and model, and the initialization of the accidental operation. Steam explosion is a hypothetical severe accident liable to happen after a core melting. The molten part of the core (called corium) falls in the lower part of the reactor. The interaction between the hot corium and the cold water remaining at the bottom of the vessel induces a massive and violent vaporization of water, similar to an explosive phenomenon. A shock wave propagates in the vessel. what can damage seriously the neighbouring structures or drill the vessel. This work presents a synthesis of in-vessel parametrical studies carried out with the Europlexus code, the linkage of the thermal-hydraulic code Mc3d dedicated to the pre-mixing phase with the Europlexus code dealing with the explosion, and finally a benchmark between the Cigalon and Europlexus codes relative to the Vulcano mock-up. (author)

  12. Thermal analysis and cooling structure design of the primary collimator in CSNS/RCS

    CERN Document Server

    Zou, Yi-Qing; Kang, Ling; Qu, Hua-Min; He, Zhe-Xi; Yu, Jie-Bing; 10.1088/1674-1137/37/5/057004

    2013-01-01

    The rapid cycling synchrotron (RCS) of the China Spallation Neutron Source (CSNS) is a high intensity proton ring with beam power of 100 kW. In order to control the residual activation to meet the requirements of hands-on maintenance, a two-stage collimation system has been designed for the RCS. The collimation system consists of one primary collimator made of thin metal to scatter the beam and four secondary collimators as absorbers. Thermal analysis is an important aspect in evaluating the reliability of the collimation system. The calculation of the temperature distribution and thermal stress of the primary collimator with different materials is carried out by using ANSYS code. In order to control the temperature rise and thermal stress of the primary collimator to a reasonable level, an air cooling structure is intended to be used. The mechanical design of the cooling structure is presented, and the cooling effciency with different chin numbers and wind velocity is also analyzed. Finally, the fatigue life...

  13. Seismic Spectrum Analysis of Advanced PWR Primary Loop and Parameter Sensitivity Study%先进压水堆一回路地震反应谱分析及参数敏感性研究

    Institute of Scientific and Technical Information of China (English)

    段蓉; 佟立丽; 曹学武

    2014-01-01

    The primary loop system of pressurized water reactor (PWR ) consists of reactor pressure vessel (RPV) ,steam generator (SG) ,reactor coolant pump (RCP) , pressurizer ,surge line and other crucial components .T he seismic response of each com-ponent is closely related to the structure of the w hole system .From a system perspec-tive ,the primary loop system of the advanced passive pressurized water reactor was studied .With ANSYS software ,the three-dimensional finite element model was built to perform modal analysis .Based on the results of modeling ,the seismic spectrum analysis with three orthotropic directions on the primary loop system was conducted to obtain the stress and displacement response .T hen sensitivity analysis of parameters ,such as spec-trum input angle and stiffness of supports was performed ,giving guidance on further design and analysis .Moreover ,direct integration method was used to get time-history response ,which was compared with spectrum simulation results .The displacement and acceleration input for seismic analysis of specific components were offered .Besides , compared with three-dimensional finite element model ,the lumped mass model of the primary loop system was also built to conduct seismic analysis ,giving the advantage and necessity of three-dimensional modeling .The study provides technical support for the structure analysis of key equipments of advanced PWR primary loop .%压水堆一回路系统包含压力容器、蒸汽发生器、主泵、稳压器、主管道和波动管等重要部件,各部件在地震激励下的动态响应与整个系统的结构形式密切相关。本文从系统的角度,以非能动先进压水堆一回路为研究对象,运用 ANSYS建立了其三维有限元模型,在模态分析的基础上,进行了三正交方向输入下的反应谱分析,得到了系统在地震载荷下的响应。并对反应谱输入角度和支撑刚度进行了敏感性研究,给出了这些特性参数

  14. Alloy 690 in PWR type reactors; Aleaciones base niquel en condiciones de primario de los reactores tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Briceno, D.; Serrano, M.

    2005-07-01

    Alloy 690, used as replacement of Alloy 600 for vessel head penetration (VHP) nozzles in PWR, coexists in the primary loop with other components of Alloy 600. Alloy 690 shows an excellent resistance to primary water stress corrosion cracking, while Alloy 600 is very susceptible to this degradation mechanisms. This article analyse comparatively the PWSCC behaviour of both Ni-based alloys and associated weld metals 52/152 and 82/182. (Author)

  15. Radiation protection performance for the dismantling of the WWR-M primary cooling circuit.

    Science.gov (United States)

    Lobach, Yu N; Luferenko, E D; Shevel, V N

    2014-12-01

    The WWR-M is a light-water-cooled and moderated heterogonous research reactor with a thermal output of 10 MW. The reactor has been in operation for >50 y and has had an excellent safety record. A non-hermeticity of the inlet line of the primary cooling circuit (PCC) was found, and the only reasonable technical solution was the complete replacement of the PCC inlet and outlet pipe lines. Such a replacement was a challenging technical task due to the necessity to handle large size components with complex geometries under conditions of high-level radiation fields, and therefore, it required detailed planning aiming to reduce staff exposure. This paper describes the dismantling and removal of the PCC components focusing on radiation protection issues.

  16. Modelling the transport of radionuclides released in the Ilha Grande bay (Brazil) after a Large Break Loca ion the primary system of a PWR

    Energy Technology Data Exchange (ETDEWEB)

    Aguiar, Andre Silva de; Simoes Filho, Francisco Fernando Lamego; Soares, Abner Duarte; Lapa, Celso Marcelo Franklin, E-mail: flamego@ien.gov.b, E-mail: asoares@cnen.gov.b, E-mail: lapa@ien.gov.b [Instituto de Engenharia Nuclear (LIMA/IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    It was postulated, in the cooling system of the core, a LOCA, where 431 m{sup 3} of soda almost instantaneously was lost. This inventory contained 1.87x10{sup 10} Bq/m{sup 3} of tritium, 2.22x10{sup 7} Bq/m{sup 3} of cobalt,3.48x10{sup 8} Bq/m{sup 3} of cesium and 3.44x10{sup 10} Bq/m{sup 3} of iodine and was released in liquid form near the Itaorna cove, Angra dos Reis - RJ. Applying the model in the proposed scenario (Angra 1 and 2 in operation and Angra 3 progressively reducing the capture and discharge after the accident), the simulated dilution of the specific activity of radionuclide spots, reached values much lower than report levels for seawater (1,1x10{sup 6} Bq/m{sup 3}, 1,11x10{sup 4} Bq/m{sup 3} and 1,85x10{sup 3} Bq/m{sup 3}) after 22 hours, respectively for {sup 3}H, {sup 60}Co, {sup 131}I and {sup 137}Cs. From the standpoint of public exposure to radionuclide dispersion, the results of activity concentration obtained by the model suggest that the observed radiological impact is negligible. Based on these findings, we conclude that there would be no radiological impact related to a further release of controlled effluent discharges into Itaorna cove. (author)

  17. Study of water radiolysis in relation with the primary cooling circuit of pressurized water reactors; Etude sur la radiolyse de l`eau en relation avec le circuit primaire de refroidissement des reacteurs nucleaires a eau sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Pastina, B

    1997-07-01

    This memorandum shows a fundamental study on the water radiolysis in relation with the cooling primary circuit of PWR type reactors. The water of the primary circuit contains boric acid a soluble neutronic poison and also hydrogen that has for role to inhibit the water decomposition under radiation effect. In the aim to better understand the mechanism of dissolved hydrogen action and to evaluate the impact of several parameters on this mechanism, aqueous solutions with boric acid and hydrogen have been irradiated in a experimental nuclear reactor, at 30, 100 and 200 Celsius degrees. It has been found that, with hydrogen, the water decomposition under irradiation is a threshold phenomenon in function of the ratio between the radiation flux `1` B(n, )`7 Li and the gamma flux. When this ratio become too high, the number of radicals is not sufficient to participate at the chain reaction, and then water is decomposed in O{sub 2} and H{sub 2}O{sub 2} in a irreversible way. The temperature has a beneficial part on this mechanism. The iron ion and the copper ion favour the water decomposition. (N.C.). 83 refs.

  18. Cooling the intact loop of primary heat transport system using Shutdown Cooling System in case of LOCA events

    Directory of Open Access Journals (Sweden)

    Icleanu Diana Laura

    2015-01-01

    Full Text Available The purpose of this paper is to model the operation of the Shutdown Cooling System (SDCS for CANDU 6 nuclear power plants in case of LOCA accidents, using Flowmaster calculation code, by delimiting models and setting calculation assumptions, input data for hydraulic analysis and input data for calculating thermal performance check for heat exchangers that are part of this system.

  19. Piping Flexibility Analysis of the Primary Cooling System of TRIGA 2000 Bandung Reactor due to Earthquake

    Directory of Open Access Journals (Sweden)

    H.P. Rahardjo

    2011-08-01

    Full Text Available Earthquakes in a nuclear installation can overload a piping system which is not flexible enough. These loads can be forces, moments and stresses working on the pipes or equipments. If the load is too large and exceed the allowable limits, the piping and equipment can be damaged and lead to overall system operation failure. The load received by piping systems can be reduced by making adequate piping flexibility, so all the loads can be transmitted homogenously throughout the pipe without load concentration at certain point. In this research the analysis of piping stress has been conducted to determine the size of loads that occured in the piping of primary cooling system of TRIGA 2000 Reactor, Bandung if an earthquake happened in the reactor site. The analysis was performed using Caesar II software-based finite element method. The ASME code B31.1 arranging the design of piping systems for power generating system (Power Piping Code was used as reference analysis method. Modeling of piping systems was based on the cooling piping that has already been installed and the existing data reported in Safety Analysis Reports (SARs of TRIGA 2000 reactor, Bandung. The quake considered in this analysis is the earthquake that occurred due to the Lembang fault, since it has the Peak Ground Acceleration (PGA in the Bandung TRIGA 2000 reactor site. The analysis results showed that in the static condition for sustain and expansion loads, the stress fraction in all piping lines does not exceed the allowable limit. However, during operation moment, in dynamic condition, the primary cooling system is less flexible at sustain load, ekspansi load, and combination load and the stress fraction have reached 95,5%. Therefore a pipeline modification (rerouting is needed to make pipe stress does not exceed the allowable stress. The pipeline modification was carried out by applied a gap of 3 mm in the X direction of the support at node 25 and eliminate the support at the node

  20. Primary Energy Efficiency Analysis of Different Separate Sensible and Latent Cooling Techniques

    Energy Technology Data Exchange (ETDEWEB)

    Abdelaziz, Omar [ORNL

    2015-01-01

    Separate Sensible and Latent cooling (SSLC) has been discussed in open literature as means to improve air conditioning system efficiency. The main benefit of SSLC is that it enables heat source optimization for the different forms of loads, sensible vs. latent, and as such maximizes the cycle efficiency. In this paper I use a thermodynamic analysis tool in order to analyse the performance of various SSLC technologies including: multi-evaporators two stage compression system, vapour compression system with heat activated desiccant dehumidification, and integrated vapour compression with desiccant dehumidification. A primary coefficient of performance is defined and used to judge the performance of the different SSLC technologies at the design conditions. Results showed the trade-off in performance for different sensible heat factor and regeneration temperatures.

  1. Primary Energy Efficiency Analysis of Different Separate Sensible and Latent Cooling Techniques

    Energy Technology Data Exchange (ETDEWEB)

    Abdelaziz, Omar [ORNL

    2015-01-01

    Separate Sensible and Latent cooling (SSLC) has been discussed in open literature as means to improve air conditioning system efficiency. The main benefit of SSLC is that it enables heat source optimization for the different forms of loads, sensible vs. latent, and as such maximizes the cycle efficiency. In this paper I use a thermodynamic analysis tool in order to analyse the performance of various SSLC technologies including: multi-evaporators two stage compression system, vapour compression system with heat activated desiccant dehumidification, and integrated vapour compression with desiccant dehumidification. A primary coefficient of performance is defined and used to judge the performance of the different SSLC technologies at the design conditions. Results showed the trade-off in performance for different sensible heat factor and regeneration temperatures.

  2. Generic study on the relation between contamination if primary coolants and occupational radiation exposure in nuclear power plants with PWR. Final report; Generische Studie zum Zusammenhang zwischen Kontamination von Primaerkreislaufmedien und beruflicher Strahlenexposition bei Kernkraftwerken mit Druckwasserreaktor. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Artmann, Andreas; Bruhn, Gerd; Schneider, Sebastian [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Koeln (Germany); Strub, Erik [Koeln Univ. (Germany)

    2016-01-15

    A generic model for the primary cooling system contamination in pressurized water reactors and the resulting radiological consequences has been developed. The functional capability was demonstrated by means of three examples concerning manipulation procedures during revision outages. Activities at the main reactor coolant pumps were studied and the influence of the coolant contamination on the resulting dose rates and collective doses were calculated. The effect of a Co-90 hot spot in a more remote area on the radiation exposure during the specific action at the reactor pumps was considered.

  3. Conceptual study on advanced PWR system

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Yoon Young; Chang, M. H.; Yu, K. J.; Lee, D. J.; Cho, B. H.; Kim, H. Y.; Yoon, J. H.; Lee, Y. J.; Kim, J. P.; Park, C. T.; Seo, J. K.; Kang, H. S.; Kim, J. I.; Kim, Y. W.; Kim, Y. H.

    1997-07-01

    In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. (1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. (2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. (3) Control rod drive mechanism for fine control : type and function were surveyed. (4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. (5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. (6) Steam injector concepts: analysis and experiment were conducted. (7) Fluidic diode concepts : analysis and experiment were conducted. (8) Wet thermal insulator : tests for thin steel layers and assessment of materials. (9) Passive residual heat removal system : a performance analysis computer code for PRHRS was developed and the conformance to EPRI requirement was checked. (author). 18 refs., 55 tabs., 137 figs.

  4. Integral Test Facility PKL: Experimental PWR Accident Investigation

    OpenAIRE

    2012-01-01

    Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR) at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circul...

  5. Effect of the primary cooling rate on the motility and fertility of frozen-thawed rabbit spermatozoa

    Directory of Open Access Journals (Sweden)

    Tatsuhiro Maeda

    2012-07-01

    Full Text Available In the present study, we examined the effect of primary cooling rates on the motility and fertility of frozen-thawed rabbit spermatozoa. Rabbit semen diluted with an egg-yolk acetamide extender was cooled from room temperature to 5°C at four different rates (-0.1, -0.2, -0.4, -0.8°C/min as a primary cooling step, then semen was frozen in liquid nitrogen vapour. After thawing, sperm cooled at -0.1°C/min showed the highest motility (40.7 ± 7.3%; there were no significant differences between the motilities of the -0.1, -0.2, and -0.4°C/min groups. The motility of frozen-thawed sperm cooled at -0.8°C/min (29.2 ± 6.8% was significantly lower than that of sperm cooled at -0.1 and -0.2°C/min. The viability (-0.1°C/min, 38.1 ± 4.0%; -0.8°C/min, 24.3 ± 7.3% of frozen-thawed sperm was closely related to its motility (-0.1°C/min, 36.7 ± 7.2%; -0.8°C/min, 22.3 ± 4.7%. Quality of post-thaw motile sperm cooled at different rates was estimated by comparing the fertilisation ability of the -0.1 and -0.8°C/min groups following artificial insemination. There were no significant differences in pregnancy rates and mean litter sizes. These data suggest that cooling rabbit semen at rates ranging from -0.1 to -0.8°C/min affects the viability but not the fertilisation capacity of motile spermatozoa after thawing.

  6. Vibration test on KMRR reactor structure and primary cooling system piping

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-10-01

    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author).

  7. Habitable Planets Around White and Brown Dwarfs: The Perils of a Cooling Primary

    CERN Document Server

    Barnes, Rory

    2012-01-01

    White and brown dwarfs are astrophysical objects that are bright enough to support an insolation habitable zone (IHZ). Unlike hydrogen-burning stars, they cool and become less luminous with time, and hence their IHZ moves in with time. The inner edge of the IHZ is defined as the orbital radius at which a planet may enter a moist or runaway greenhouse, phenomena that can remove a planet's surface water forever. Thus, as the IHZ moves in, planets that enter it may no longer have any water, and are still uninhabitable. Additionally, the close proximity of the IHZ to the primary leads to concern that tidal heating may also be strong enough to trigger a runaway greenhouse, even for orbital eccentricities as small as 10^-6. Water loss occurs due to photolyzation by UV photons in the planetary stratosphere, followed by hydrogen escape. Young white dwarfs emit a large amount of these photons as their surface temperatures are over 10^4 K. The situation is less clear for brown dwarfs, as observational data do not const...

  8. Habitable planets around white and brown dwarfs: the perils of a cooling primary.

    Science.gov (United States)

    Barnes, Rory; Heller, René

    2013-03-01

    White and brown dwarfs are astrophysical objects that are bright enough to support an insolation habitable zone (IHZ). Unlike hydrogen-burning stars, they cool and become less luminous with time; hence their IHZ moves in with time. The inner edge of the IHZ is defined as the orbital radius at which a planet may enter a moist or runaway greenhouse, phenomena that can remove a planet's surface water forever. Thus, as the IHZ moves in, planets that enter it may no longer have any water and are still uninhabitable. Additionally, the close proximity of the IHZ to the primary leads to concern that tidal heating may also be strong enough to trigger a runaway greenhouse, even for orbital eccentricities as small as 10(-6). Water loss occurs due to photolyzation by UV photons in the planetary stratosphere, followed by hydrogen escape. Young white dwarfs emit a large amount of these photons, as their surface temperatures are over 10(4) K. The situation is less clear for brown dwarfs, as observational data do not constrain their early activity and UV emission very well. Nonetheless, both types of planets are at risk of never achieving habitable conditions, but planets orbiting white dwarfs may be less likely to sustain life than those orbiting brown dwarfs. We consider the future habitability of the planet candidates KOI 55.01 and 55.02 in these terms and find they are unlikely to become habitable.

  9. 百万级压水堆核电站空冷汽轮机选型研究%Calculation and Selecting of Air-cooled Turbine in 1 000 MW PWR Nuclear Power Station

    Institute of Scientific and Technical Information of China (English)

    杨建军; 孙传军; 赵迪

    2015-01-01

    针对国内某核电厂址的初步气象条件,进行了百万等级核电机组空冷汽轮机选型计算分析。分析结果表明,该厂址参数下若选用半转速核电汽轮机,目前已有的末级叶片不能满足要求,需要开发全新的空冷末级叶片;而利用大型火电空冷汽轮机已有运行业绩的末级叶片,开发全转速百万级核电空冷汽轮机,将花费较少的时间和经费,是比较适宜的方案。某厂址气象条件在我国北方具有代表性,上述工作为我国北方缺水地区核电站汽轮机选型提供了参考。建议国内汽轮机厂利用已有核电、火电汽轮机技术积累,尽快启动全转速百万级核电空冷汽轮机研发工作,形成具有完全自主知识产权的百万级核电空冷汽轮机技术,抢占技术制高点。%By analyzing the weather characters of one potential 1000MW nuclear power station site ,this paper calculated the air-cooled steam turbine exhausting area .It is concluded that the present last long blade used in half-speed turbine is not suitable,new last long blade shall be developed if we want to use half-speed turbine in 1 000MW air-cooled nuclear power station.However,the full-speed turbines and their last long blades used in coal burning power station are appropriate for this site,excepting plus one low pressure turbine and some measures to remove moisture steam in the last stage .The latter is better,for it will cost less money and time .The site mentioned in this paper is representive in China north area ,and this paper is useful for turbine selecting of nuclear power station in this area .Domestic turbine manufacturers have accumulated many experiences and technologies by manufacturing and developing half -speed nuclear turbine and air-cooled full-speed turbines in coal burning power station .It is suggested the domestic turbine manufacturers should setup the research work as soon as possible .

  10. Estimating probable flaw distributions in PWR steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Gorman, J.A.; Turner, A.P.L. [Dominion Engineering, Inc., McLean, VA (United States)

    1997-02-01

    This paper describes methods for estimating the number and size distributions of flaws of various types in PWR steam generator tubes. These estimates are needed when calculating the probable primary to secondary leakage through steam generator tubes under postulated accidents such as severe core accidents and steam line breaks. The paper describes methods for two types of predictions: (1) the numbers of tubes with detectable flaws of various types as a function of time, and (2) the distributions in size of these flaws. Results are provided for hypothetical severely affected, moderately affected and lightly affected units. Discussion is provided regarding uncertainties and assumptions in the data and analyses.

  11. The Possibility of Building Nuclear Power Plant Free from Severe Accident Risk PWR NPP with advanced all passive safety cooling systems (AAP SCS)%发展无严重事故风险核电站的曙光具有完全非能动安全冷却系统的压水堆核电站

    Institute of Scientific and Technical Information of China (English)

    肖宏才

    2013-01-01

    A complete set of advanced all passive safety cooling systems (AAP SCS) for PWR NPP,actuated by natural force has been put forward in the article.Here the natural force mainly means the fore,which created by change of pressure distribution in the first loop of PWR as a result of operational regime conversion from one to another,including occurrence of accident situation.Correspondent safety cooling system will be actuated naturally and then put it into passive operation after occurring some kind of accident,so accidental situation will be mitigated right after it's occurrence and core residual heat will be naturally moved from the active core to the ultimate heat sink.There is no need to rely on automatic control system,any active equipment and human actions in all working process of the AAP SCS,which can reduce the probability of severe accident to zero,so as to exclude the need of evacuation plan around AAP nuclear power plant and eliminate the public's concern and doubt about nuclear power safety.Implementation of the AAP SCS concept is only based on use of evolutionary measures and state-of-the-art technology.So at present time it can be used for design of new-type third generation PWR nuclear power plant without severe accident risk,and for modernization of existing second generation nuclear power plant.%本文提出了用自然力直接触发启动压水堆核电站一整套完全非能动的停堆安全冷却系统.这里的自然力主要是指一回路运行工况转换时由于其压力分布变化所形成的压差力.在这一系统中,当进行停堆或发生某种一回路事故工况时,相应的安全冷却系统便自然地投入运行,立即缓解事故后果,将事故时一回路释放的能量及堆芯余热非能动地排入最终热阱.在全过程中不依靠自动控制系统、能动设备及任何人为因素的介入,即可确保对堆芯余热无限期的安全冷却能力,完全避免压水堆核电站发生向环境泄漏放射性物

  12. RELAP5 Analyses of ROSA/LSTF Experiments on AM Measures during PWR Vessel Bottom Small-Break LOCAs with Gas Inflow

    Directory of Open Access Journals (Sweden)

    Takeshi Takeda

    2014-01-01

    Full Text Available RELAP5 code posttest analyses were performed on ROSA/LSTF experiments that simulated PWR 0.2% vessel bottom small-break loss-of-coolant accidents with different accident management (AM measures under assumptions of noncondensable gas inflow and total failure of high-pressure injection system. Depressurization of and auxiliary feedwater (AFW injection into the secondary-side of both steam generators (SGs as the AM measures were taken 10 min after a safety injection signal. The primary depressurization rate of 55 K/h caused rather slow primary depressurization being obstructed by the gas accumulation in the SG U-tubes after the completion of accumulator coolant injection. Core temperature excursion thus took place by core boil-off before the actuation of low-pressure injection (LPI system. The fast primary depressurization by fully opening the relief valves in both SGs with continuous AFW injection led to long-term core cooling by the LPI actuation even under the gas accumulation in the SG U-tubes. The code indicated remaining problems in the predictions of break flow rate during two-phase flow discharge period and primary pressure after the gas inflow. Influences of the primary depressurization rate with continuous AFW injection onto the long-term core cooling were clarified by the sensitivity analyses.

  13. CIRCUMFERENTIAL INHOMOGENITY ANALYSIS IN G.A. SIWABESSY REACTOR’S PRIMARY COOLING PIPE

    Directory of Open Access Journals (Sweden)

    Roziq Himawan

    2016-10-01

    Full Text Available ABSTRACT In the in-service inspection conducted to G.A. Siwabessy reactor’s primary cooling system pipe, it was found the presence of inhomogenity inside of welding part. To verify whether the inhomogenity could be tolerated or not, comparative data from welding pre-service inspection is needed. Unfortunately, this weld wasn’t covered in pre-service inspection. Therefore, this inhomogenity needs to be analyzed. The purpose of this study is to evaluate the stress intensity factor of the inhomogenity, whether it is within a limit value or not and to predict the crack growth. Analysis were performed based on fracture mechanics theory using parameter of stress intensity factor. Two models were used for calculation approach that are plane crack model and semi-elliptic crack model. Hence, in order to predict the length of inhomogenity in the future, crack growth calculations were performed. The results showed that stress intensity values from both two models are remain below fracture toughness value of pipe’s material. Besides that, stress intensity factor from plane crack model is higher than those from semi-elliptic crack model. Under consideration that inhomogenity has an arc shape in actual, thus, stress intensity factor from this inhomogenity still low enough compare to the fracture toughness. Crack growth calculation’s results showed that after 300th cycle of loading, the length of inhomogenity reaches approximately 2 mm. Based on operation data of G.A. Siwabessy reactor, 300 cycle number is corresponds to 30 years operation. Based on these results it could be concluded that the presence of inhomogenity in the welding part does not affect the structure’s integrity of piping system. Keywords : Inhomogenity, fracture mechanics, fracture toughness, stress intensity factor, crack growth   ABSTRAK Pada pelaksanaan in-service inspection terhadap perpipaan sistem pendingin primer reaktor G.A. Siwabessy diketahui adanya inhomogenitas pada

  14. SCOR 1000: an economic and innovative conceptual design PWR

    Energy Technology Data Exchange (ETDEWEB)

    Gautier, G.M.; Chenaud, M.S. [CEA Cadarache (DEN/DER/SESI), 13 - Saint Paul lez Durance (France). Dept. d' Etudes des Reacteurs; Tourniaire, B. [CEA Grenoble (DEN/DTN/SE2T/LPTM), 38 (France)

    2007-07-01

    Within the framework of innovative reactors studies, the Cea proposes the SCOR design (Simple COmpact Reactor) based on most of the advantages of innovative reactors. All main components are integrated in the vessel: the pressurizer, the canned pumps, the control rod mechanics of the driving system (CMD), and the dedicated heat exchangers of the passive heat removal system. The only steam generator is located above the vessel instead of the upper head. This design is featured by its compactness and by a large suppression or simplification of auxiliary systems. The first design with a 600 MWe shows its competitiveness with regard to the large loop-type PWR. To reduce the cost investment by the law sized effect, we examine the possibility of increasing the power of the reactor, while keeping the safety advantages of the medium sized SCOR. The electrical power of the new design is 1000 MWe. SCOR-1000 operates at much lower primary circuit pressure than standard PWRs (93 bars instead of the usual 155 bars), and the power density is lower (80 MW/m3 instead of 100 for the present PWRs). The reactivity is controlled by the CMD and by the burnable poison, without soluble boron. With the same safety advantages of the medium-sized SCOR, the cost reduction of the investment and of cost production could reach 18% with regard to the loop-type PWR. (authors)

  15. Neutron noise measurements on Bugey 2 PWR

    Energy Technology Data Exchange (ETDEWEB)

    Marini, J.; Romy, D.; Spadi, J.C.; Assedo, R.; Castello, G.

    1982-01-01

    Following Bugey 2 PWR hot functional tests, dimension measurements of internals hold down spring led to suspect that vibration levels could change with time. Neutron noise measurements runs during the first cycle enabled describing vibration behaviour of internals. Comparisons with previous analytical and experimental results gained on the Safran model as well as on similar reactors were also made.

  16. Differential roles of galanin on mechanical and cooling responses at the primary afferent nociceptor

    Directory of Open Access Journals (Sweden)

    Hulse Richard P

    2012-06-01

    Full Text Available Abstract Background Galanin is expressed in a small percentage of intact small diameter sensory neurons of the dorsal root ganglia and in the afferent terminals of the superficial lamina of the dorsal horn of the spinal cord. The neuropeptide modulates nociception demonstrating dose-dependent pro- and anti-nociceptive actions in the naïve animal. Galanin also plays an important role in chronic pain, with the anti-nociceptive actions enhanced in rodent neuropathic pain models. In this study we compared the role played by galanin and its receptors in mechanical and cold allodynia by identifying individual rat C-fibre nociceptors and characterising their responses to mechanical or acetone stimulation. Results Mechanically evoked responses in C-fibre nociceptors from naive rats were sensitised after close intra-arterial infusion of galanin or Gal2-11 (a galanin receptor-2/3 agonist confirming previous data that galanin modulates nociception via activation of GalR2. In contrast, the same dose and route of administration of galanin, but not Gal2-11, inhibited acetone and menthol cooling evoked responses, demonstrating that this inhibitory mechanism is not mediated by activation of GalR2. We then used the partial saphenous nerve ligation injury model of neuropathic pain (PSNI and the complete Freund’s adjuvant model of inflammation in the rat and demonstrated that close intra-arterial infusion of galanin, but not Gal2-11, reduced cooling evoked nociceptor activity and cooling allodynia in both paradigms, whilst galanin and Gal2-11 both decreased mechanical activation thresholds. A previously described transgenic mouse line which inducibly over-expresses galanin (Gal-OE after nerve injury was then used to investigate whether manipulating the levels of endogenous galanin also modulates cooling evoked nociceptive behaviours after PSNI. Acetone withdrawal behaviours in naive mice showed no differences between Gal-OE and wildtype (WT mice. 7-days after

  17. DNBR Analysis of a Dual-Cooled Annular Fuel for the OPR1000 Application

    Energy Technology Data Exchange (ETDEWEB)

    Shin, C. H.; In, W. K.; Oh, D. S.; Chun, T. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    A dual-cooled annular fuel for a pressurized water reactor (PWR) has been introduced for a significant amount of reactor power uprate. The Korea Atomic Energy Research Institute (KAERI) has been performing a research to develop a dual-cooled annular fuel for the power uprate of 20% in an optimized PWR in Korea, OPR1000. Several thermal-hydraulic tasks exist for the application of the dual-cooled annular fuel to OPR1000. The primary task is the balance of the minimum DNBR (MDNBR) between the inner and outer channels since the coolant flows through the circular inner channel of annular fuel as well as the outer subchannels formed between the fuel rods. The MDNBR balance has been known to largely depend on the thermal conductance in the inner and outer gaps. Another task is to evaluate the operating condition that the inner coolant channel is partially blocked. This study calculated the MDNBR in the inner and outer channels depending on the thermal gap conductance, i.e., inner and outer gap width. The acceptable range of gap width is determined for the MDNBR not to exceed the DNBR limit during anticipated operational occurrences (AOOs) as well as normal operation. The limit for the flow blockage in the inner channel is also estimated based on the DNBR analysis

  18. PWR reactor vessel in-service-inspection according to RSEM

    Energy Technology Data Exchange (ETDEWEB)

    Algarotti, Marc; Dubois, Philippe; Hernandez, Luc; Landez, Jean Paul [Intercontrole, 13, rue du Capricorne - SILIC 433, 94583 Rungis - Cedex (France)

    2006-07-01

    Nuclear services experience Framatome ANP (an AREVA and Siemens company) has designed and constructed 86 Pressurized Water Reactors (PWR) around the world including the three units lately commissioned at Ling Ao in the People's Republic of China and ANGRA 2 in Brazil; the company provided general and specialized outage services supporting numerous outages. Along with the American and German subsidiaries, Framatome ANP Inc. and Framatome ANP GmbH, Framatome ANP is among the world leading nuclear services providers, having experience of over 500 PWR outages on 4 continents, with current involvement in more than 50 PWR outages per year. Framatome ANP's experience in the examinations of reactor components began in the 1970's. Since then, each unit (American, French and German companies) developed automated NDT inspection systems and carried out pre-service and ISI (In-Service Inspections) using a large range of NDT techniques to comply with each utility expectations. These techniques have been validated by the utilities and the safety authorities of the countries where they were implemented. Notably Framatome ANP is fully qualified to provide full scope ISI services to satisfy ASME Section XI requirements, through automated NDE tasks including nozzle inspections, reactor vessel head inspections, steam generator inspections, pressurizer inspections and RPV (Reactor Pressure Vessel) inspections. Intercontrole (Framatome ANP subsidiary dedicated in supporting ISI) is one of the leading NDT companies in the world. Its main activity is devoted to the inspection of the reactor primary circuit in French and foreign PWR Nuclear Power Plants: the reactor vessel, the steam generators, the pressurizer, the reactor internals and reactor coolant system piping. NDT methods mastered by Intercontrole range from ultrasonic testing to eddy current and gamma ray examinations, as well as dye penetrant testing, acoustic monitoring and leak testing. To comply with the high

  19. Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

    Institute of Scientific and Technical Information of China (English)

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

    High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (~22.5%), increase the energy output (~41%), decrease the quantity of spent fuels to be disposed (~2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

  20. Calculation of source term in spent PWR fuel assemblies for dry storage and shipping cask design; Calculo de los terminos fuente de combustibles irradiados PWR para el diseno de contenedores de almacenamiento y transporte

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, J. L.; Lopez, J.

    1986-07-01

    Using the ORIGEN-2 Coda, the decay heat and neutron and photon sources for an irradiated PWR fuel element have been calculated. Also, parametric studies on the behaviour of the magnitudes with the burn-up, linear heat power and irradiation and cooling times were performed. Finally, a comparison between our results and other design calculations shows a good agreement and confirms the validity of the used method. (Author) 6 refs.

  1. Shielding design for PWR in France

    Energy Technology Data Exchange (ETDEWEB)

    Champion, G.; Charransol; Le Dieu de Ville, A.; Nimal, J.C.; Vergnaud, T.

    1983-05-01

    Shielding calculation scheme used in France for PWR is presented here for 900 MWe and 1300 MWe plants built by EDF the French utility giving electricity. Neutron dose rate at areas accessible by personnel during the reactor operation is calculated and compared with the measurements which were carried out in 900 MWe units up to now. Measurements on the first French 1300 MWe reactor are foreseen at the end of 1983.

  2. The integrated PWR; Les REP integres

    Energy Technology Data Exchange (ETDEWEB)

    Gautier, G.M. [CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d' Etudes des Reacteurs

    2002-07-01

    This document presents the integrated reactors concepts by a presentation of four reactors: PIUS, SIR, IRIS and CAREM. The core conception, the operating, the safety, the economical aspects and the possible users are detailed. From the performance of the classical integrated PWR, the necessity of new innovative fuels utilization, the research of a simplified design to make easier the safety and the KWh cost decrease, a new integrated reactor is presented: SCAR 600. (A.L.B.)

  3. Modeling and Simulation of Release of Radiation in Flow Blockage Accident for Two Loops PWR

    OpenAIRE

    Khurram Mehboob; Cao Xinrong; Majid Ali

    2012-01-01

    In this study modeling and simulation of release of radiation form two loops PWR has been carried out for flow blockage accident. For this purpose, a MATLAB based program “Source Term Evaluator for Flow Blockage Accident” (STEFBA) has been developed, which uses the core inventory as its primary input. The TMI-2 reactor is considered as the reference plant for this study. For 1100 reactor operation days, the core inventory has been evaluated under the core design constrains at average reactor ...

  4. Analysis of N-16 concentration in primary cooling system of AP1000 power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rohanda, Anis [Center for Reactor Technology and Nuclear Safety – BATAN Kawasan PUSPIPTEK Gd. No. 80 Serpong, Tangerang Selatan 15310 (Indonesia); Waris, Abdul [Physics Department of ITB, Indonesia anis-rohanda@yahoo.com (Indonesia)

    2015-04-16

    Nitrogen-16 (N-16) is one of the radiation safety parameter on the primary reactor system. The activation product, N-16, is the predominant contributor to the activity in the reactor coolant system during reactor operation. N-16 is activation product derived from activation of O-16 with fast neutron based on {sup 16}O(n,p){sup 16}N reaction. Thus study is needed and it performs to determine N-16 concentration in reactor coolant (primary coolant) in supporting radiation safety. One of the way is using analytical methode based on activation and redecay princip to obtain N-16 concentration. The analysis was performed on the configuration basis and operational of Westinghouse AP1000 power reactor in several monitoring points at coolant reactor system. The results of the calculation of N-16 concentration at the core outlet, reactor vessel outlet, pressurizer line, inlet and outlet of steam generators, primary pumps, reactor vessels inlet and core inlet are: 281, 257, 255, 250, 145, 142, 129 and 112 µCi/gram respectively. The results of analysis compared with AP1000 design control document as standard values. The verification showed very high accuracy comparation between analytical results and standard values.

  5. Retention of PWR primary coolant trace elements by cation exchange resins during cold shutdown with oxygenation: modelling and experimental results for silver behavior; Retention des elements traces du fluide primaire des REP par les resines echangeuses de cations lors des mises en arret a froid avec oxygenation: modelisation et resultats experimentaux relatifs au comportement de l'argent

    Energy Technology Data Exchange (ETDEWEB)

    Elain, L.; Doury-Berthod, M. [CEA Saclay, INSTN, Institut National des Sciences et Techniques Nucleaires, 91 - Gif-sur-Yvette (France); Genin, J.B. [CEA Cadarache, Dir. de l' Energie Nucleaire (DEN), 13 - Saint-Paul-lez-Durance (France); Berger, M. [Electricite de France (EDF/SEPTEN), 69 - Villeurbanne (France)

    2004-07-01

    In order to minimize the radiochemical impact of the corrosion products on the operation of Pressurized Water Reactors, on-line purification of the primary coolant is carried out. The purification system arranged on the Chemical and Volume Control System is made up of mechanical filters and demineralizers packed with a mixed bed of cation and anion exchange resins. This paper proposes an update on the retention of primary coolant trace elements by the cation exchange resins of the demineralizers during cold shutdowns with oxygenation. The study is first of all devoted to the description of the concentration profiles of the various cation constituents which settle in the demineralizer during purification after oxygenation. For a number of trace elements, localized enrichment zones at the Li{sup +}/Ni(Il) exchange zone are expected to appear in the column. The case of silver is afterwards discussed in detail. Thermodynamic modelling shows that the theoretical retention volume of the metallic element and its degree of enrichment in the column are dependent on the basic composition of the primary coolant and the specific characteristics of the demineralizer cation exchanger. At the Ag{sup +} ion concentration expected in the reactor coolant after oxygenation (between 10{sup -8} mol.L{sup -1} and 10{sup -6} mol.L{sup -1}), the breakthrough of silver should be near-simultaneous with that of nickel. The experimental results, obtained in the laboratory and with a 'Mini-CVCS' pilot instrumentation recently used during the cold shutdown of Tricastin Unit 2,900 MWe PWR NPP, confirm the validity of these theoretical forecasts and enable new hypotheses to be advanced for explaining silver release from a demineralizer. (authors)

  6. Effects of Lower Drying-Storage Temperature on the Ductility of High-Burnup PWR Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Burtseva, T. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-08-30

    The purpose of this research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures (PCT) and hoop stresses. Extensive precipitation of radial hydrides could lower the failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods stored in pools, at temperatures below the ductile-to-brittle transition temperature (DBTT).

  7. Aqueous Nanofluid as a Two-Phase Coolant for PWR

    Directory of Open Access Journals (Sweden)

    Pavel N. Alekseev

    2012-01-01

    Full Text Available Density fluctuations in liquid water consist of two topological kinds of instant molecular clusters. The dense ones have helical hydrogen bonds and the nondense ones are tetrahedral clusters with ice-like hydrogen bonds of water molecules. Helical ordering of protons in the dense water clusters can participate in coherent vibrations. The ramified interface of such incompatible structural elements induces clustering impurities in any aqueous solution. These additives can enhance a heat transfer of water as a two-phase coolant for PWR due to natural forming of nanoparticles with a thermal conductivity higher than water. The aqueous nanofluid as a new condensed matter has a great potential for cooling applications. It is a mixture of liquid water and dispersed phase of extremely fine quasi-solid particles usually less than 50 nm in size with the high thermal conductivity. An alternative approach is the formation of gaseous (oxygen or hydrogen nanoparticles in density fluctuations of water. It is possible to obtain stable nanobubbles that can considerably exceed the molecular solubility of oxygen (hydrogen in water. Such a nanofluid can convert the liquid water in the nonstoichiometric state and change its reduction-oxidation (RedOx potential similarly to adding oxidants (or antioxidants for applying 2D water chemistry to aqueous coolant.

  8. Study of colloidal particles behaviour in the PWR primary circuit conditions; Etude du comportement des particules colloidales dans les conditions physicochimiques du circuit primaire des reacteurs a eau sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Barale, M

    2006-12-15

    EDF wants to understand, model and limit primary circuit contamination of Pressurized Water Reactors by colloidal particles resulting from corrosion. The electrostatic behaviour of representative oxide particles (cobalt ferrite, nickel ferrite and magnetite) has been studied in primary circuit conditions with the influence of boric acid and lithium hydroxide. The isoelectric point (IEP) and the point of zero charge (PZC) of particles, measured between 5 C and 320 C, exhibit a minimum towards 200 C. The thermodynamic constants of the protonation equilibrium of surface sites were calculated. When boric acid is added, zeta potential and IEP decrease because of borate ions sorption. On the contrary, there is not effect of lithium ions. The modelling of these results under conditions representative of primary circuit shows that these oxides exhibit a negative surface charge, explaining their sorption and adhesion behaviour. (author)

  9. Electrochemistry of Water-Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

    2006-08-08

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

  10. Horizontal Drop of 21- PWR Waste Package

    Energy Technology Data Exchange (ETDEWEB)

    A.K. Scheider

    2001-04-26

    The objective of this calculation is to determine the structural response of the waste package (WP) dropped horizontally from a specified height. The WP used for that purpose is the 21-Pressurized Water Reactor (PWR) WP. The scope of this document is limited to reporting the calculation results in terms of stress intensities. The information provided by the sketches (Attachment I) is that of the potential design of the type of WP considered in this calculation, and all obtained results are valid for that design only. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 16). AP-3.12Q, ''Calculations'' (Ref. 11) is used to perform the calculation and develop the document. The sketches attached to this calculation provide the potential dimensions and materials for the 21-PWR WP design.

  11. Effect of transplutonium doping on approach to long-life core in uranium-fueled PWR

    Energy Technology Data Exchange (ETDEWEB)

    Peryoga, Yoga; Saito, Masaki; Artisyuk, Vladimir [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors; Shmelev, Anatolii [Moscow Engineering Physics Institute, Moscow (Russian Federation)

    2002-08-01

    The present paper advertises doping of transplutonium isotopes as an essential measure to improve proliferation-resistance properties and burnup characteristics of UOX fuel for PWR. Among them {sup 241}Am might play the decisive role of burnable absorber to reduce the initial reactivity excess while the short-lived nuclides {sup 242}Cm and {sup 244}Cm decay into even plutonium isotopes, thus increasing the extent of denaturation for primary fissile {sup 239}Pu in the course of reactor operation. The doping composition corresponds to one discharged from a current PWR. For definiteness, the case identity is ascribed to atomic percentage of {sup 241}Am, and then the other transplutonium nuclide contents follow their ratio as in the PWR discharged fuel. The case of 1 at% doping to 20% enriched uranium oxide fuel shows the potential of achieving the burnup value of 100 GWd/tHM with about 20% {sup 238}Pu fraction at the end of irradiation. Since so far, americium and curium do not require special proliferation resistance measures, their doping to UOX would assist in introducing nuclear technology in developing countries with simultaneous reduction of accumulated minor actinides stockpiles. (author)

  12. Study of safety relief valve operation under ATWS conditions. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Hutmacher, E.S.; Nesmith, B.J.; Brukiewa, J.B.

    1979-06-25

    A literature survey and analysis project has been performed to determine if recent (since mid-1975) data has been reported which could influence the current approach to predicting PWR relief valve capacity under ATWS conditions. This study was conducted by the Energy Technology Engineering Center for NRC. Results indicate that the current relief valve capacity model tends to predict less capacity than actually obtains; however, no experimental verification at PWR ATWS conditions was found. Other project objectives were to establish the availability of methods for evaluating reaction forces and back pressure effects on relief valve capacity, and to determine if facilities exist which are capable of testing PWR relief valves at ATWS conditions.

  13. Integral Test Facility PKL: Experimental PWR Accident Investigation

    Directory of Open Access Journals (Sweden)

    Klaus Umminger

    2012-01-01

    Full Text Available Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circulation pumps and steam generators (SGs arranged symmetrically around the reactor pressure vessel (RPV. The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermal-hydraulic phenomena. This paper presents a survey of test objectives and programs carried out to date. It also describes the test facility in its present state. Some important results obtained over the years with focus on investigations carried out since the beginning of the international cooperation are exemplarily discussed.

  14. Procedure qualification of CNP650 PWR primary coolant pipeline by manual welding%CNP650型压水堆主管道手工焊接工艺评定

    Institute of Scientific and Technical Information of China (English)

    刘先文

    2012-01-01

    The primary coolant pipe of CNP650 pressurized water reactor is the enterclose of coolant of reactor core,which is the pressure pipe of large diameter and thickness connected with RPV(reactor pressure vessel) and SG(steam generator) and RCP(reactor coolant pump).The welding construction of primary coolant pipe is the pivotal path of the installation of main equipment of nuclear island and the key and difficult point of the construction of nuclear power plant.The data sheet and welding experience of the WPQ is very important for ensuring the success of the first welding construction.The process control of the manual WPQ of CNP650 nuclear power plant of QinShan Nuclear Power Phase II Expansion Project included the simulation condition of the site and the management of welding process and physical and chemical testing and welding deformation, in order to get the deposited metal fitting for the requirements of the NDE and physical and chemical properties of the technical specification.The process control is the prerequisite of the welding construction of primary coolant pipe.%CNP650型压水堆的主管道作为反应堆压力容器堆芯冷却剂的通道,是连接反应堆压力容器、主泵和蒸汽发生器的大型厚壁承压管道.主管道焊接施工是核岛主设备安装的关键路径,是核电建设的重点与难点.焊接工艺评定所提供的数据与焊接经验,对确保主管道焊接施工一次成功,起着非常重要的作用.泰山核电二期扩建工程CNP650型核电站主管道手工焊接工艺评定从模拟现场焊接施工的条件、焊接过程管理、理化试验、焊接变形等方面进行控制,以获得符合技术规范对熔敷金属无损检测、理化性能的要求.焊接工艺评定过程控制为主管道焊接施工提供先决条件.

  15. Effect of co-free valve on activity reduction in PWR

    Energy Technology Data Exchange (ETDEWEB)

    Bahn, C.B.; Han, B.C.; Bum, J.S.; Hwang, I.S. [Department of Nuclear Engineering, Seoul National Univ. (Korea, Republic of); Lee, C.B. [Korea Atomic Energy Research Inst., Daejon (Korea, Republic of)

    2002-07-01

    Radioactive nuclei, such as {sup 68}Co and {sup 60}Co, deposited on out-of-core surfaces in a pressurized water reactor (PWR) primary coolant system, are major sources of occupational radiation exposure to plant maintenance personnel and act as costly impediment to prompt and effective repairs. Valve hardfacing alloys exposed to primary coolant are considered as one of the main Co sources. To evaluate the Co-free valve, such as NOREM 02 and Deloro 50, the candidates for the alternative to Stellite 6, in a simulated PWR primary condition, SNU corrosion test loop (SCOTL) was constructed. For gate valves hard-faced with made of NOREM 02 and Deloro 50 hot cycling tests were conducted for up to 2,000 on-off cycles with cold leak tests at 1,000 cycle interval. It was observed that the leak rate of NOREM 02 (Fe-base) did not satisfy the nuclear grade valve leak criteria. After 1000 cycles test, while there was no leakage in case of Deloro 50 (Ni-base). Also, Deloro 50 showed no leakage after 2000 cycles. To estimate the activity reduction effect, we modified CRUDSIM-MIT which modeled the effects of coolant chemistry on the crud transport and activity buildup in the primary system of PWR. In the new code, crud evaluation and assessment (CREAT), {sup 60}Co activity buildup prediction includes 1) Co-base valve replacement effect, 2) Co-base valve maintenance effect, and 3) control rod drive mechanism (CRDM) and main coolant pump (MCP) shaft contribution. CREAT predicted that the main contributor of Co activity buildup was the corrosion-induced release of Co from the steam generator (SG) tubing. With new SG's tubed with alloy 690, Korean Next Generation Reactor (APR-1400) is expected to have about 64% lower Co activity on SG surface. The use of all Co-free valves is expected to cut additional 8% of activity which is only marginal. (authors)

  16. Characterization of Factors affecting IASCC of PWR Core Internals

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Woo; Hwang, Seong Sik; Kim, Won Sam [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-09-15

    A lot works have been performed on IASCC in BWR. Recent efforts have been devoted to investigate IASCC in PWR, but the mechanism in PWR is not fully understood yet as compared with that in BWR due to a lack of data from laboratories and fields. Therefore it is strongly needed to review and analyse recent researches of IASCC in both BWR and PWR for establishing a proactive management technology for IASCC of core internals in Korean PWRs. This work is aimed to review mainly recent technical reports on IASCC of stainless steels for core internals in PWR. For comparison, the works on IASCC in BWR were also reviewed and briefly introduced in this report.

  17. The PWR cores management; La gestion des coeurs REP

    Energy Technology Data Exchange (ETDEWEB)

    Barral, J.C. [Electricite de France (EDF), 75 - Paris (France); Rippert, D. [CEA Cadarache, Departement d' Etudes des Reacteurs, DER, 13 - Saint-Paul-lez-Durance (France); Johner, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee, DRFC, 13 - Saint-Paul-lez-Durance (France)] [and others

    2000-01-25

    During the meeting of the 25 january 2000, organized by the SFEN, scientists and plant operators in the domain of the PWR debated on the PWR cores management. The five first papers propose general and economic information on the PWR and also the fast neutron reactors chains in the electric power market: statistics on the electric power industry, nuclear plant unit management, the ITER project and the future of the thermonuclear fusion, the treasurer's and chairman's reports. A second part offers more technical papers concerning the PWR cores management: performance and optimization, in service load planning, the cores management in the other countries, impacts on the research and development programs. (A.L.B.)

  18. Zebra: An advanced PWR lattice code

    Energy Technology Data Exchange (ETDEWEB)

    Cao, L.; Wu, H.; Zheng, Y. [School of Nuclear Science and Technology, Xi' an Jiaotong Univ., No. 28, Xianning West Road, Xi' an, ShannXi, 710049 (China)

    2012-07-01

    This paper presents an overview of an advanced PWR lattice code ZEBRA developed at NECP laboratory in Xi'an Jiaotong Univ.. The multi-group cross-section library is generated from the ENDF/B-VII library by NJOY and the 361-group SHEM structure is employed. The resonance calculation module is developed based on sub-group method. The transport solver is Auto-MOC code, which is a self-developed code based on the Method of Characteristic and the customization of AutoCAD software. The whole code is well organized in a modular software structure. Some numerical results during the validation of the code demonstrate that this code has a good precision and a high efficiency. (authors)

  19. Degraded core analysis for the PWR

    Energy Technology Data Exchange (ETDEWEB)

    Gittus, J.H.

    1987-10-01

    The paper presents an analysis of the probability and consequences of degraded core accidents for the PWR. The article is based on a paper which was presented by the author to the Sizewell-B public inquiry. Degraded core accidents are examined with respect to:- the initiating events, safety plant failure, and processes with a bearing on containment failure. Accident types and frequencies are discussed, as well as the dispersion of radionuclides. Accident risks, i.e. individual and societal risks in degraded core accidents are assessed from:- the amount of radionuclides released, the weather, the population distribution, and the accident frequencies. Uncertainties in the assessment of degraded core accidents are also summarized. (U.K.).

  20. A pressure drop model for PWR grids

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dong Seok; In, Wang Ki; Bang, Je Geon; Jung, Youn Ho; Chun, Tae Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A pressure drop model for the PWR grids with and without mixing device is proposed at single phase based on the fluid mechanistic approach. Total pressure loss is expressed in additive way for form and frictional losses. The general friction factor correlations and form drag coefficients available in the open literatures are used to the model. As the results, the model shows better predictions than the existing ones for the non-mixing grids, and reasonable agreements with the available experimental data for mixing grids. Therefore it is concluded that the proposed model for pressure drop can provide sufficiently good approximation for grid optimization and design calculation in advanced grid development. 7 refs., 3 figs., 3 tabs. (Author)

  1. Coupled Analysis on Steady Flow in Primary and Secondary Sides of PWR Steam Generator%压水堆蒸汽发生器一、二次侧稳态流场耦合分析

    Institute of Scientific and Technical Information of China (English)

    丛腾龙; 田文喜; 秋穗正; 苏光辉; 谢永诚; 姚彦贵

    2014-01-01

    The steam generator (SG) suffers from the challenge of tube rupture caused by the flow induced vibration (FIV ) during operation .T he 3D flow characteristics in SG are essential for the analysis of FIV .The secondary side flow field was simulated based on the porous media model ,using the coupled heat transfer from primary to secondary side . T he 3D velocity , temperature , pressure and quality distributions in secondary side ,the one-dimensional temperature and heat transfer coefficient (HTC) distributions of primary and secondary sides as well as the U-tube temperature distribution were obtained .The thermal-hydraulic characteristic distributions in the shell side and the flow vapor quality distribution in separators are significantly uneven due to the non-uniformly distributed heat source released from primary to secondary side .The flow quality distribution in the separators varies from 0.05 to 0.62 ,w hich can be used to the design for separator load .The velocity distribution in the U-bend region was calculated , w hich provides input conditions for the evaluation of FIV damage of tubes .%蒸汽发生器(S G )在运行过程中主要面临流致振动所导致的传热管破裂事故,而流致振动分析需以SG内的三维两相流场作为输入条件。采用多孔介质模型,对SG二次侧流场进行求解,同时耦合一、二次侧换热,获得SG二次侧速度场、温度场、压力场及流动含气率分布,并获得传热管一维的一、二次侧流体温度和换热系数及传热管温度分布。由于一次侧向二次侧释热极不均匀,SG内流场分布及汽水分离器内的含气率分布极不均匀;汽水分离器内的最大、最小含气率分别为0.62和0.05,该参数可为汽水分离器负载设计提供依据。通过计算还获得弯管区速度分布,该分布可为传热管的流致振动磨损评估提供输入条件。

  2. VERIFIKASI KECELAKAAN HILANGNYA ALIRAN AIR UMPAN PADA REAKTOR DAYA PWR MAJU

    Directory of Open Access Journals (Sweden)

    Andi Sofrany Ekariansyah

    2015-03-01

    Reactor Technology and Nuclear Safety as a Technical Support Organization (TSO in terms of reactor safety verification, the verification activities have been carried out for the AP1000 that begins with failure of secondary coolant accident verification. The activity started with the technical safety features modeling such as passive core cooling system consisting of a Passive Residual Heat Removal system (PRHR, Core Makeup Tank (CMT, and In-containment Refueling Water Storage Tank (IRWST. The failure of secondary coolant accident selected is the loss of main feedwater flow to one of the steam generator simulated using the calculation program RELAP5/SCDAP/Mod3.4. The objective of analysis is to obtain sequences of changes in the thermalhydraulic parameters in the reactor due to the selected event. Analysis results obtained are validated and compared with the analysis results using the calculation program LOFTRAN in the AP1000 safety design document. The verification results show that the loss of feed-water supply has no impact on core damage, the reactor coolant system, as well as secondary systems. The ability of heat exchanger PRHR has been verified to dissipate decay heat of the core after reactor trip. Validation with the AP1000 safety design document shows compliance on most thermal hydraulic parameters. In general, the advanced PWR model equipped with passive core cooling system that has been developed remains safe in the event of loss of secondary coolant flow accident. Keywords: Verification, loss of feed water flow, AP1000

  3. Component testing of a ground based gas turbine steam cooled rich-burn primary zone combustor for emissions control of nitrogeneous fuels

    Science.gov (United States)

    Schultz, D. F.

    1986-01-01

    This effort summarizes the work performed on a steam cooled, rich-burn primary zone, variable geometry combustor designed for combustion of nitrogeneous fuels such as heavy oils or synthetic crude oils. The steam cooling was employed to determine its feasibility and assess its usefulness as part of a ground based gas turbine bottoming cycle. Variable combustor geometry was employed to demonstrate its ability to control primary and secondary zone equivalence ratios and overall pressure drop. Both concepts proved to be highly successful in achieving their desired objectives. The steam cooling reduced peak liner temperatures to less than 800 K. This low temperature offers the potential of both long life and reduced use of strategic materials for liner fabrication. These degrees of variable geometry were successfully employed to control air flow distribution within the combustor. A variable blade angle axial flow air swirler was used to control primary zone air flow, while the secondary and tertiary zone air flows were controlled by rotating bands which regulated air flow to the secondary zone quench holes and the dilutions holes respectively.

  4. PWR circuit contamination assessment tool. Use of OSCAR code for engineering studies at EDF

    Directory of Open Access Journals (Sweden)

    Benfarah Moez

    2016-01-01

    Full Text Available Normal operation of PWR generates corrosion and wear products in the primary circuit which are activated in the core and constitute the major source of the radiation field. In addition, cases of fuel failure and alpha emitter dissemination in the coolant system could represent a significant radiological risk. Radiation field and alpha risks are the main constraints to carry out maintenance and to handle effluents. To minimize these risks and constraints, it is essential to understand the behavior of corrosion products and actinides and to carry out the appropriate measurements in PWR circuits and loop experiments. As a matter of fact, it is more than necessary to develop and use a reactor contamination assessment code in order to take into account the chemical and physical mechanisms in different situations in operating reactors or at design stage. OSCAR code has actually been developed and used for this aim. It is presented in this paper, as well as its use in the engineering studies at EDF. To begin with, the code structure is described, including the physical, chemical and transport phenomena considered for the simulation of the mechanisms regarding PWR contamination. Then, the use of OSCAR is illustrated with two examples from our engineering studies. The first example of OSCAR engineering studies is linked to the behavior of the activated corrosion products. The selected example carefully explores the impact of the restart conditions following a reactor mid-cycle shutdown on circuit contamination. The second example of OSCAR use concerns fission products and disseminated fissile material behavior in the primary coolant. This example is a parametric study of the correlation between the quantity of disseminated fuel and the variation of Iodine 134 in the primary coolant.

  5. PWR safety/relief valve blowdown analysis experience

    Energy Technology Data Exchange (ETDEWEB)

    Lee, M.Z.; Chou, L.Y.; Yang, S.H. (Gilbert/Commonwealth Engineers and Consultants, Reading, PA (USA). Speciality Engineering Dept.)

    1982-10-01

    The paper describes the difficulties encountered in analyzing a PWR primary loop pressurizer safety relief valve and power operated relief valve discharge system, as well as their resolution. The experience is based on the use of RELAP5/MOD1 and TPIPE computer programs as the tools for fluid transient analysis and piping dynamic analysis, respectively. General approaches for generating forcing functions from thermal fluid analysis solution to be used in the dynamic analysis of piping are reviewed. The paper demonstrates that the 'acceleration or wave force' method may have numerical difficulties leading to unrealistic, large amplitude, highly oscillatory forcing functions in the vicinity of severe flow area discontinuities or choking junctions when low temperature loop seal water is discharged. To avoid this problem, an alternate computational method based on the direct force method may be used. The simplicity and superiority in numerical stability of the forcing function computation method as well as its drawbacks are discussed. Additionally, RELAP modeling for piping, valve, reducer, and sparger is discussed. The effects of loop seal temperature on SRV and PORV discharge line blowdown forces, pressure and temperature distributions are examined. Finally, the effects of including support stiffness and support eccentricity in piping analysis models, method and modeling relief tank connections, minimization of tank nozzle loads, use of damping factors, and selection of solution time steps are discussed.

  6. Application of LBB to high energy piping systems in operating PWR

    Energy Technology Data Exchange (ETDEWEB)

    Swamy, S.A.; Bhowmick, D.C. [Westinghouse Nuclear Technology Division, Pittsburgh, PA (United States)

    1997-04-01

    The amendment to General Design Criterion 4 allows exclusion, from the design basis, of dynamic effects associated with high energy pipe rupture by application of leak-before-break (LBB) technology. This new approach has resulted in substantial financial savings to utilities when applied to the Pressurized Water Reactor (PWR) primary loop piping and auxiliary piping systems made of stainless steel material. To date majority of applications pertain to piping systems in operating plants. Various steps of evaluation associated with the LBB application to an operating plant are described in this paper.

  7. The continued development of the MFM suite and its practical application on a PWR system

    DEFF Research Database (Denmark)

    Thunem, Harald P-J; Zhang, Xinxin

    2015-01-01

    This paper reports on the results from the practical application of the Shape Shifter framework on the continued development of a graphical editing suite, the MFM Suite, for MFM and process model design and analysis. The primary use of the MFM Suite is diagnosis and prognosis of anomalies...... in physical processes. One of the Halden Reactor Project’s advanced NPP simulators based on a PWR is used to demonstrate the applicability of the suite in realistic situations. The paper presents a summary and suggests some plans for future research and development....

  8. Thermal analysis of a storage cask for 24 spent PWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J.C.; Bang, K.S.; Seo, K.S.; Kim, H.D. [Korea Atomic Energy Research Inst., Daejeon (Korea); Choi, B.I.; Lee, H.Y.; Song, M.J. [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea)

    2004-07-01

    The purpose of this paper is to perform a thermal analysis of a spent fuel storage cask in order to predict the maximum concrete and fuel cladding temperatures. Thermal analyses have been carried out for a storage cask under normal and off-normal conditions. The environmental temperature is assumed to be 27 {open_square} under the normal condition. The off-normal condition has an environmental temperature of 40 {open_square}. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Four of the eight inlet ducts are assumed to be completely blocked. The storage cask is designed to store 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat load from the 24 PWR assemblies is 25.2 kW. Thermal analyses of ventilation system have been carried out for the determination of the optimum duct size and shape. The finite volume computational fluid dynamics code FLUENT was used for the thermal analysis. In the results of the analysis, the maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal condition and off-normal conditions.

  9. Experiment data report for semiscale Mod-1 Test S-06-5. (LOFT counterpart test). [PWR

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-06-01

    Recorded test data are presented for Test S-06-5 of the Semiscale Mod-1 LOFT counterpart test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-06-5 was conducted from initial conditions of 2272 psia and 536/sup 0/F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold legs of the intact and broken loops to simulate emergency core coolant injection in a PWR. The purpose of Test S-06-5 was to assess the influence of the break nozzle geometry on core thermal and system response and on the subcooled and low quality mass flow rates at the break locations.

  10. Conceptual study of advanced PWR core design. Development of advanced PWR core neutronics analysis system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hyo; Kim, Seung Cho; Kim, Taek Kyum; Cho, Jin Young; Lee, Hyun Cheol; Lee, Jung Hun; Jung, Gu Young [Seoul National University, Seoul (Korea, Republic of)

    1995-08-01

    The neutronics design system of the advanced PWR consists of (i) hexagonal cell and fuel assembly code for generation of homogenized few-group cross sections and (ii) global core neutronics analysis code for computations of steady-state pin-wise or assembly-wise core power distribution, core reactivity with fuel burnup, control rod worth and reactivity coefficients, transient core power, etc.. The major research target of the first year is to establish the numerical method and solution of multi-group diffusion equations for neutronics code development. Specifically, the following studies are planned; (i) Formulation of various numerical methods such as finite element method(FEM), analytical nodal method(ANM), analytic function expansion nodal(AFEN) method, polynomial expansion nodal(PEN) method that can be applicable for the hexagonal core geometry. (ii) Comparative evaluation of the numerical effectiveness of these methods based on numerical solutions to various hexagonal core neutronics benchmark problems. Results are follows: (i) Formulation of numerical solutions to multi-group diffusion equations based on numerical methods. (ii) Numerical computations by above methods for the hexagonal neutronics benchmark problems such as -VVER-1000 Problem Without Reflector -VVER-440 Problem I With Reflector -Modified IAEA PWR Problem Without Reflector -Modified IAEA PWR Problem With Reflector -ANL Large Heavy Water Reactor Problem -Small HTGR Problem -VVER-440 Problem II With Reactor (iii) Comparative evaluation on the numerical effectiveness of various numerical methods. (iv) Development of HEXFEM code, a multi-dimensional hexagonal core neutronics analysis code based on FEM. In the target year of this research, the spatial neutronics analysis code for hexagonal core geometry(called NEMSNAP-H temporarily) will be completed. Combination of NEMSNAP-H with hexagonal cell and assembly code will then equip us with hexagonal core neutronics design system. (Abstract Truncated)

  11. Application of a PID controller based on fuzzy logic to reduce variations in the control parameters in PWR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Vasconcelos, Wagner Eustaquio de; Lira, Carlos Alberto Brayner de Oliveira; Brito, Thiago Souza Pereira de; Afonso, Antonio Claudio Marques, E-mail: wagner@unicap.br, E-mail: cabol@ufpe.br, E-mail: afonsofisica@gmail.com, E-mail: thiago.brito86@yahoo.com.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Centro de Tecnologia e Geociencias. Departamento de Energia Nuclear; Cruz Filho, Antonio Jose da; Marques, Jose Antonio, E-mail: antonio.jscf@gmail.com, E-mail: jamarkss@uol.com.br [Universidade Catolica de Pernambuco (CCT/PUC-PE), Recife, PE (Brazil). Centro de Ciencias e Tecnologia; Teixeira, Marcello Goulart, E-mail: marcellogt@dcc.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), Rio de Janeiro, RJ (Brazil). Instituto de Matematica. Dept. de Matematica

    2013-07-01

    Nuclear reactors are in nature nonlinear systems and their parameters vary with time as a function of power level. These characteristics must be considered if large power variations occur in power plant operational regimes, such as in load-following conditions. A PWR reactor has a component called pressurizer, whose function is to supply the necessary high pressure for its operation and to contain pressure variations in the primary cooling system. The use of control systems capable of reducing fast variations of the operation variables and to maintain the stability of this system is of fundamental importance. The best-known controllers used in industrial control processes are proportional-integral-derivative (PID) controllers due to their simple structure and robust performance in a wide range of operating conditions. However, designing a fuzzy controller is seen to be a much less difficult task. Once a Fuzzy Logic controller is designed for a particular set of parameters of the nonlinear element, it yields satisfactory performance for a range of these parameters. The objective of this work is to develop fuzzy proportional-integral-derivative (fuzzy-PID) control strategies to control the level of water in the reactor. In the study of the pressurizer, several computer codes are used to simulate its dynamic behavior. At the fuzzy-PID control strategy, the fuzzy logic controller is exploited to extend the finite sets of PID gains to the possible combinations of PID gains in stable region. Thus the fuzzy logic controller tunes the gain of PID controller to adapt the model with changes in the water level of reactor. The simulation results showed a favorable performance with the use to fuzzy-PID controllers. (author)

  12. Open and closed fuel cycle of HWR and PWR. How large is the high-level radioactive wastes repository; Ciclos de combustible abierto y cerrado con HWR y PWR. ?Cuanto mas grande es el repositorio de residuos radiactivos de alta actividad?

    Energy Technology Data Exchange (ETDEWEB)

    Bevilacqua, Arturo M. [Comision Nacional de Energia Atomica, San Carlos de Bariloche (Argentina). Centro Atomico Bariloche

    1996-07-01

    A conceptual analysis was carried out on the size of a high-level wastes (HLW) repository for the waste arising from once-through and closed fuel cycles with (HLW) and PWR. The mass, the activity and thermal loading was calculated with the ORIGEN2.1 computer code for the spent fuel and for the high-level liquid wastes. It was considered a minimum burnup of 7.000 MW.d/t U and 33.000 MW.d/t U for HWR and PWR respectively, cooling times of 20 and 55 years, reprocessing recovery ratios of 99% and 99.7% and a total electricity production of 81.6 GW(e).a. It was concluded that the cooling time is the most important repository size reproduction parameter for the closed cycles. On the other hand, the spent fuel mass for the once-through cycles does not depend on the cooling time what prevents repository size reduction once a cooling time of 55 years is reached. The repository size reduction in the case of HWR is larger than in the case of PWR, owing to the larger fuel mass required to produce the specific electricity amount. (author)

  13. A PWR Thorium Pin Cell Burnup Benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, Kevan Dean; Zhao, X.; Pilat, E. E; Hejzlar, P.

    2000-05-01

    As part of work to evaluate the potential benefits of using thorium in LWR fuel, a thorium fueled benchmark comparison was made in this study between state-of-the-art codes, MOCUP (MCNP4B + ORIGEN2), and CASMO-4 for burnup calculations. The MOCUP runs were done individually at MIT and INEEL, using the same model but with some differences in techniques and cross section libraries. Eigenvalue and isotope concentrations were compared on a PWR pin cell model up to high burnup. The eigenvalue comparison as a function of burnup is good: the maximum difference is within 2% and the average absolute difference less than 1%. The isotope concentration comparisons are better than a set of MOX fuel benchmarks and comparable to a set of uranium fuel benchmarks reported in the literature. The actinide and fission product data sources used in the MOCUP burnup calculations for a typical thorium fuel are documented. Reasons for code vs code differences are analyzed and discussed.

  14. Conceptual study of advanced PWR core design

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong

    1997-09-01

    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs.

  15. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyu-Tae, E-mail: ktkim@dongguk.ac.kr

    2013-10-15

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10{sup −6} on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure.

  16. Seismic qualification of PWR plant auxiliary feedwater systems

    Energy Technology Data Exchange (ETDEWEB)

    Lu, S.C.; Tsai, N.C.

    1983-08-01

    The NRC Standard Review Plan specifies that the auxiliary feedwater (AFW) system of a pressurized water reactor (PWR) is a safeguard system that functions in the event of a Safe Shutdown Earthquake (SSE) to remove the decay heat via the steam generator. Only recently licensed PWR plants have an AFW system designed to the current Standard Review Plan specifications. The NRC devised the Multiplant Action Plan C-14 in order to make a survey of the seismic capability of the AFW systems of operating PWR plants. The purpose of this survey is to enable the NRC to make decisions regarding the need of requiring the licensees to upgrade the AFW systems to an SSE level of seismic capability. To implement the first phase of the C-14 plan, the NRC issued a Generic Letter (GL) 81-14 to all operating PWR licensees requesting information on the seismic capability of their AFW systems. This report summarizes Lawrence Livermore National Laboratory's efforts to assist the NRC in evaluating the status of seismic qualification of the AFW systems in 40 PWR plants, by reviewing the licensees' responses to GL 81-14.

  17. Fatigue Crack Growth Rate of Type 347 Stainless Steel at the PWR Environment

    Energy Technology Data Exchange (ETDEWEB)

    Min, Ki Deuk; Kim, Seon Jin [Hanyang University, Seoul (Korea, Republic of); Kim, Dae Whan; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Materials used in nuclear power plants are low alloy steel, stainless steel, and superalloy steel. Understanding the characteristics of these materials is important in the development of nuclear power plant related technology. Nb-stabilized Type 347 stainless steel is used for the coolant pressurizer surge line of Korea Standard Nuclear Power Plant (KSNPP). Surge line of PWR nuclear reactor are damaged by thermal fatigue due to thermal gradient during heat-up and cool-down, mechanical fatigue due to mechanical stress, and corrosion fatigue due to nuclear reactor water environment. Fatigue is an important factor which limits the life of structure. Fatigue crack growth rate curves in nuclear reactor environment are needed to evaluate the integrity of nuclear reactor structure but that result is not sufficient. In this study, fatigue crack growth rates at nuclear reactor environment are produced to evaluate integrity of nuclear power plant section 5

  18. The advanced main control console for next japanese PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Tsuchiya, A. [Hokkaido Electric Power Co., Inc., Sapporo (Japan); Ito, K. [Mitsubishi Heavy Industries, Ltd., Nuclear Energy Systems Engineering Center, Yokohama (Japan); Yokoyama, M. [Mitsubishi Electric Corporation, Energy and Industrial Systems Center, Kobe (Japan)

    2001-07-01

    The purpose of the improvement of main control room designing in a nuclear power plant is to reduce operators' workload and potential human errors by offering a better working environment where operators can maximize their abilities. In order to satisfy such requirements, the design of main control board applied to Japanese Pressurized Water Reactor (PWR) type nuclear power plant has been continuously modified and improved. the Japanese Pressurized Water Reactor (PWR) Utilities (Electric Power Companies) and Mitsubishi Group have developed an advanced main control board (console) reflecting on the study of human factors, as well as using a state of the art electronics technology. In this report, we would like to introduce the configuration and features of the Advanced Main Control Console for the practical application to the next generation PWR type nuclear power plants including TOMARI No.3 Unit of Hokkaido Electric Power Co., Inc. (author)

  19. Primary volcanic structures from a type section of Deccan Trap flows around Narsingpur–Harrai–Amarwara, central India: Implications for cooling history

    Indian Academy of Sciences (India)

    Piyali Sengupta; Arijit Ray

    2006-12-01

    Field investigations of the Deccan Trap lava sequence along a 70 km traverse in the Narsingpur–Harrai–Amarwara area of central India indicate twenty lava flows comprising a total thickness of around 480m. Primary volcanic structures like vesicles and cooling joints are conspicuous in this volcanic succession and are used to divide individual flows into three well-defined zones namely the lower colonnade zone, entablature zone, and the upper colonnade zone. The variable nature of these structural zones is used for identification and correlation of lava flows in the field. For twenty lava flows, the thicknesses of upper colonnade zones of eight flows are ∼5m while those of eight other flows are ∼8m each. The thicknesses of upper colonnade zones of remaining four flows could not be measured in the field. Using the thicknesses of these upper colonnade zones and standard temperature-flow thickness-cooling time profiles for lava pile, the total cooling time of these sixteen Deccan Trap lava flows has been estimated at 12 to 15 years.

  20. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Unit Nuclear Energy, Netherlands Energy Research Foundation ECN, Petten (Netherlands)); Hoogenboorm, J.E.; De Leege, P.F.A. (International Reactor Institute IRI, University of Leiden, Leiden (Netherlands)); Van de Voet, J.; Verhagen, F.C.M. (KEMA NV, Arnhem (Netherlands))

    1992-01-01

    In order to carry out reliable reactor core calculations for a boiled water reactor (BWR) or a pressurized water reactor (PWR) first reactivity calculations have to be carried out for which several calculation programs are available. The purpose of the title project is to exchange experiences to improve the knowledge of this reactivity calculations. In a large number of institutes reactivity calculations of PWR and BWR pin cells were executed by means of available computer codes. Results are compared. It is concluded that the variations in the calculated results are problem dependent. Part of the results is satisfactory. However, further research is necessary.

  1. Monte Carlo based radial shield design of typical PWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

    2016-11-15

    Neutron and gamma flux and dose equivalent rate distribution are analysed in radial and shields of a typical PWR type reactor based on the Monte Carlo radiation transport computer code MCNP5. The ENDF/B-VI continuous energy cross-section library has been employed for the criticality and shielding analysis. The computed results are in good agreement with the reference results (maximum difference is less than 56 %). It implies that MCNP5 a good tool for accurate prediction of neutron and gamma flux and dose rates in radial shield around the core of PWR type reactors.

  2. Advanced ion exchange resins for PWR condensate polishing

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, B. [Rohm and Haas Co. (United States); Tsuzuki, S. [Rohm and Haas Co. (Japan)

    2002-07-01

    The severe chemical and mechanical requirements of a pressurized water reactor (PWR) condensate polishing plant (CPP) present a major challenge to the design of ion exchange resins. This paper describes the development and initial operating experience of improved cation and anion exchange resins that were specifically designed to meet PWR CPP needs. Although this paper focuses specifically on the ion exchange resins and their role in plant performance, it is also recognized and acknowledged that excellent mechanical design and operation of the CPP system are equally essential to obtaining good results. (authors)

  3. Study of the distribution of hydrogen in a PWR containment with CFD codes; Estudio de la distribucion de hidrogeno en una contencion PWR con codigos CFD

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, G.; Martinez, R. M.; Fernandez, K.; Morato, D. J.; Bocanegra Melian, R.; Mena, L.; Queral, C.

    2014-07-01

    During the development of a severe accident in a PWR reactor can be generated, large quantities of hydrogen by the oxidation of metals present in the nucleus, mainly the zirconium fuel pods. This hydrogen, along with steam and other gases, can be released to the atmosphere of contention by a leak or break in the primary circuit and achieving conditions in which combustion may occur. Combustion causes thermal and pressure loads that can damage the security systems and the integrity of the containment building, last barrier of confinement of radioactive materials. The main condition that defines the characteristics of the combustion is the concentration of species, detailed knowledge of the distribution of hydrogen is very important to correctly predict the possible damage to the containment in the event that there is combustion. (Author)

  4. Thermal hydraulic investigations and optimization on the EVC system of a PWR by CFD simulation

    Energy Technology Data Exchange (ETDEWEB)

    Xi, Mengmeng [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China); Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China); Tang, Mao [China Nuclear Power Design Engineering Co., Ltd., 518124 Shenzhen (China); Wang, Chenglong; Zheng, Meiyin; Qiu, Suizheng [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China)

    2015-08-15

    Highlights: • This study constructs a full CFD model for the EVC system of a PWR. • The complex fluid and solid coupling is treated in the computation. • Primary characteristics of the velocity, pressure and temperature distributions in the EVC system are investigated. • The optimization of the EVC system with different inlet boundaries are performed. - Abstract: In order to optimize the design of Reactor Pit Ventilation (EVC) system in a Pressurized Water Reactor (PWR), it is necessary to study the characteristics of the velocity, pressure and temperature fields in the EVC system. A full computational fluid dynamics (CFD) model for the EVC system is constructed by a commercial CFD code, where the complex fluid and solid coupling is treated. The Shear Stress Transport (SST) model is adopted to perform the turbulence calculation. This paper numerically investigates the characteristics of the velocity, pressure and temperature distributions in the EVC system. In particular, the effects of inlet air parameters on the thermal hydraulic characteristics and the reactor pit structure are also discussed for the EVC system optimization. Simulations are carried out with different mesh sizes and boundary conditions for sensitivity analysis. The computational results are important references to optimize the design and verify the rationality of the EVC system.

  5. PWR Containment Shielding Calculations with SCALE6.1 Using Hybrid Deterministic-Stochastic Methodology

    Directory of Open Access Journals (Sweden)

    Mario Matijević

    2016-01-01

    Full Text Available The capabilities of the SCALE6.1/MAVRIC hybrid shielding methodology (CADIS and FW-CADIS were demonstrated when applied to a realistic deep penetration Monte Carlo (MC shielding problem of a full-scale PWR containment model. Automatic preparation of variance reduction (VR parameters is based on deterministic transport theory (SN method providing the space-energy importance function. The aim of this paper was to determine the neutron-gamma dose rate distributions over large portions of PWR containment with uniformly small MC uncertainties. The sources of ionizing radiation included fission neutrons and photons from the reactor and photons from the activated primary coolant. We investigated benefits and differences of FW-CADIS over CADIS methodology for the objective of the uniform MC particle density in the desired tally regions. Memory intense deterministic module was used with broad group library “v7_27n19g” opposed to the fine group library “v7_200n47g” used for final MC simulation. Compared with CADIS and with the analog MC, FW-CADIS drastically improved MC dose rate distributions. Modern shielding problems with large spatial domains require not only extensive computational resources but also understanding of the underlying physics and numerical interdependence between SN-MC modules. The results of the dose rates throughout the containment are presented and discussed for different volumetric adjoint sources.

  6. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

    1991-12-01

    Benchmark results of the Dutch PINK working group on PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs., 9 figs., 30 tabs.

  7. Studies of a small PWR for onsite industrial power

    Energy Technology Data Exchange (ETDEWEB)

    Klepper, O.H.; Smith, W.R.

    1977-04-19

    Information on the use of a 300 to 400 MW(t) PWR type reactor for industrial applications is presented concerning the potential market, reliability considerations, reactor plant description, construction techniques, comparison between nuclear and fossil-fired process steam costs, alternative fossil-fired steam supplies, and industrial application.

  8. PWR fuel in Japan; The changes and trend for hereafter

    Energy Technology Data Exchange (ETDEWEB)

    Yokote, Mitsuhiro (Kansai Electric Power Co., Inc., Osaka (Japan)); Kondo, Yoshiaki; Abeta, Sadaaki

    1992-07-01

    As for the PWR fuel in Japan, much efforts have been exerted aiming at the high reliability since the start of operation of Mihama No. 1 plant of Kansai Electric Power Co., Inc. At the beginning of 1970s, the fuel made by Westinghouse in USA was imported, and since then, the pursuit of the causes of troubles and the countermeasures and the domestic production of fuel have been carried out, and the improvement of design and the strengthening of quality control have been advanced. As the results, the occurrence of troubles decreased rapidly. As the fuel improvement for hereafter, the economical improvement by higher burnup, the saving and effective use of uranium resources as well as the increase of reliability are emphasized. The changes in the PWR fuel by Westinghouse, the course of improvement in the PWR fuel in Japan, the improvement against the troubles of the fuel, the improved design, the verification of the performance of the PWR fuel, the trend of development of the fuel such as the heightening of burnup, the saving and effective use of uranium resources, and the improved type pressurized water reactors are reported. (K.I.).

  9. A neutronic study of the cycle PWR-CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alberto da; Pereira, Claubia; Veloso, Maria Auxiliadora Fortini; Fortini, Angela; Pinheiro, Ricardo Brant [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear]. E-mail: albertomoc@terra.com.br; claubia@nuclear.ufmg.br; dora@nuclear.ufmg.br; fortini@nuclear.ufmg.br; rbp@nuclear.ufmg.br

    2007-07-01

    The cycle PWR-CANDU was simulated using the WIMSD-5B and ORIGEN2.1 codes. It was simulated a fuel burnup of 33,000 MWd/t for UO{sub 2} with enrichment of 3.2% and a fuel extended burnup of 45,000 MWd/t for UO{sub 2} with enrichments of 3.5%, 4.0% and 5.0% in a PWR reactor. The PWR discharged fuel was submitted to the simulation of deposition for five years. After that, it was submitted to AYROX reprocessing and used to produce a fuel to CANDU reactor. Then, it was simulated the burnup in the CANDU. Parameters such as infinite medium multiplication factor, k{sub inf}, fuel temperature coefficient of reactivity, {alpha}{sub TF}, moderator temperature coefficient of reactivity, {alpha}{sub TM}, the ratio rapid flux/total flux and the isotopic composition in the begin and the end of life were evaluated. The results showed that the fuels analyzed could be used on PWR and CANDU reactors without the need of change on the design of these reactors. (author)

  10. Methodology for the LABIHS PWR simulator modernization

    Energy Technology Data Exchange (ETDEWEB)

    Jaime, Guilherme D.G.; Oliveira, Mauro V., E-mail: gdjaime@ien.gov.b, E-mail: mvitor@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The Human-System Interface Laboratory (LABIHS) simulator is composed by a set of advanced hardware and software components whose goal is to simulate the main characteristics of a Pressured Water Reactor (PWR). This simulator serves for a set of purposes, such as: control room modernization projects; designing of operator aiding systems; providing technological expertise for graphical user interfaces (GUIs) designing; control rooms and interfaces evaluations considering both ergonomics and human factors aspects; interaction analysis between operators and the various systems operated by them; and human reliability analysis in scenarios considering simulated accidents and normal operation. The simulator runs in a PA-RISC architecture server (HPC3700), developed nearby 2000's, using the HP-UX operating system. All mathematical modeling components were written using the HP Fortran-77 programming language with a shared memory to exchange data from/to all simulator modules. Although this hardware/software framework has been discontinued in 2008, with costumer support ceasing in 2013, it is still used to run and operate the simulator. Due to the fact that the simulator is based on an obsolete and proprietary appliance, the laboratory is subject to efficiency and availability issues, such as: downtime caused by hardware failures; inability to run experiments on modern and well known architectures; and lack of choice of running multiple simulation instances simultaneously. This way, there is a need for a proposal and implementation of solutions so that: the simulator can be ported to the Linux operating system, running on the x86 instruction set architecture (i.e. personal computers); we can simultaneously run multiple instances of the simulator; and the operator terminals run remotely. This paper deals with the design stage of the simulator modernization, in which it is performed a thorough inspection of the hardware and software currently in operation. Our goal is to

  11. Effect of dissolved oxygen content on stress corrosion cracking of a cold worked 316L stainless steel in simulated pressurized water reactor primary water environment

    Science.gov (United States)

    Zhang, Litao; Wang, Jianqiu

    2014-03-01

    Stress corrosion crack growth tests of a cold worked nuclear grade 316L stainless steel were conducted in simulated pressurized water reactor (PWR) primary water environment containing various dissolved oxygen (DO) contents but no dissolved hydrogen. The crack growth rate (CGR) increased with increasing DO content in the simulated PWR primary water. The fracture surface exhibited typical intergranular stress corrosion cracking (IGSCC) characteristics.

  12. The effects of cold rolling orientation and water chemistry on stress corrosion cracking behavior of 316L stainless steel in simulated PWR water environments

    Science.gov (United States)

    Chen, Junjie; Lu, Zhanpeng; Xiao, Qian; Ru, Xiangkun; Han, Guangdong; Chen, Zhen; Zhou, Bangxin; Shoji, Tetsuo

    2016-04-01

    Stress corrosion cracking behaviors of one-directionally cold rolled 316L stainless steel specimens in T-L and L-T orientations were investigated in hydrogenated and deaerated PWR primary water environments at 310 °C. Transgranular cracking was observed during the in situ pre-cracking procedure and the crack growth rate was almost not affected by the specimen orientation. Locally intergranular stress corrosion cracks were found on the fracture surfaces of specimens in the hydrogenated PWR water. Extensive intergranular stress corrosion cracks were found on the fracture surfaces of specimens in deaerated PWR water. More extensive cracks were found in specimen T-L orientation with a higher crack growth rate than that in the specimen L-T orientation with a lower crack growth rate. Crack branching phenomenon found in specimen L-T orientation in deaerated PWR water was synergistically affected by the applied stress direction as well as the preferential oxidation path along the elongated grain boundaries, and the latter was dominant.

  13. Fatigue Crack Growth Rate Behavior of Type 347 Stainless Steel in Simulated PWR Water Environment

    Energy Technology Data Exchange (ETDEWEB)

    Min, Ki Deuk; Kim, Seon Jin [Hanyang University, Seoul (Korea, Republic of); Kim, Dae Whan; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    The pressurizer surge line of a Korean standard nuclear power plane uses Nb stabilized type 347 stainless steel. The pressurizer surge line is the pipe connecting the pressurizer and the hot leg line, and the path controlling the pressure and temperature of the cooling system of the nuclear reactor, operated at 316 .deg. C and in a 150atm. The pressurizer surge line operated at high temperature and high pressure receives thermal stress by a temperature change and mechanical stress by a pressure change at the same time, and by being exposed to the high temperature and high pressure cooling water environment of a nuclear power plant, environmental fatigue by stress and corrosion is the main damage instrument. As the effect of environmental fatigue has been reported, through low cycle fatigue, fatigue life evaluations of austenite stainless steel have been conducted, but evaluations of fatigue crack growth rate to evaluate the soundness are very poor. In this study, evaluated characteristics of fatigue crack growth rate base on a change of dissolved oxygen in a PWR environment

  14. PWR type reactors. Normal and accidental operation; Reacteurs a eau sous pression. Fonctionnement normal et accidentel

    Energy Technology Data Exchange (ETDEWEB)

    Petetrot, J.F. [AREVA NP, Dept. Fonctionnement Reacteur et Etudes d' Accidents/Division, Tour AREVA, 92 - Paris La Defense (France)

    2009-07-15

    This article presents the general operation principles of PWR type reactors with the limits to be respected for the core and the steam supply system. Regulation systems controlling the main parameters are described as well: measurements used, functional structures, controlled actuators. The protection system which can lead to the automatic shutdown of the reactor (emergency rod drop) and to the start-up of safeguard functions is detailed. The interface for the conventional protection system is briefly described. The operation of the steam supply system with respect to the power grid needs is presented in relation with the regulation of the turbogenerator set: load follow-up, primary and secondary adjustment. Finally, the changes of the most important parameters during typical transients are commented. The main operations needed to move from the cold shutdown state to the nominal power are described as well. (J.S.)

  15. Methodology of a PWR containment analysis during a thermal-hydraulic accident

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Dayane F.; Sabundjian, Gaiane; Lima, Ana Cecilia S., E-mail: dayane.silva@usp.br, E-mail: gdjian@ipen.br, E-mail: aclima@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The aim of this work is to present the methodology of calculation to Angra 2 reactor containment during accidents of the type Loss of Coolant Accident (LOCA). This study will be possible to ensure the safety of the population of the surroundings upon the occurrence of accidents. One of the programs used to analyze containment of a nuclear plant is the CONTAIN. This computer code is an analysis tool used for predicting the physical conditions and distributions of radionuclides inside a containment building following the release of material from the primary system in a light-water reactor during an accident. The containment of the type PWR plant is a concrete building covered internally by metallic material and has limits of design pressure. The methodology of containment analysis must estimate the limits of pressure during a LOCA. The boundary conditions for the simulation are obtained from RELAP5 code. (author)

  16. Representing Operational Knowledge of PWR Plant by Using Multilevel Flow Modelling

    DEFF Research Database (Denmark)

    Zhang, Xinxin; Lind, Morten; Jørgensen, Sten Bay

    2014-01-01

    situation and support operational decisions. This paper will provide a general MFM model of the primary side in a standard Westinghouse Pressurized Water Reactor ( PWR ) system including sub - systems of Reactor Coolant System, Rod Control System, Chemical and Volume Control System, emergency heat removal......The aim of this paper is to explore the capability of representing operational knowledge by using Multilevel Flow Modelling ( MFM ) methodology. The paper demonstrate s how the operational knowledge can be inserted into the MFM models and be used to evaluate the plant state, identify the current...... systems. And the sub - systems’ functions will be decomposed into sub - models according to different operational situations. An operational model will be developed based on the operating procedure by using MFM symbols and this model can be used to implement coordination rules for organize the utilizati...

  17. Electron cooling

    Science.gov (United States)

    Meshkov, I.; Sidorin, A.

    2004-10-01

    The brief review of the most significant and interesting achievements in electron cooling method, which took place during last two years, is presented. The description of the electron cooling facilities-storage rings and traps being in operation or under development-is given. The applications of the electron cooling method are considered. The following modern fields of the method development are discussed: crystalline beam formation, expansion into middle and high energy electron cooling (the Fermilab Recycler Electron Cooler, the BNL cooler-recuperator, cooling with circulating electron beam, the GSI project), electron cooling in traps, antihydrogen generation, electron cooling of positrons (the LEPTA project).

  18. Assessment of PWR plutonium burners for nuclear energy centers

    Energy Technology Data Exchange (ETDEWEB)

    Frankel, A J; Shapiro, N L

    1976-06-01

    The purpose of the study was to explore the performance and safety characteristics of PWR plutonium burners, to identify modifications to current PWR designs to enhance plutonium utilization, to study the problems of deploying plutonium burners at Nuclear Energy Centers, and to assess current industrial capability of the design and licensing of such reactors. A plutonium burner is defined to be a reactor which utilizes plutonium as the sole fissile addition to the natural or depleted uranium which comprises the greater part of the fuel mass. The results of the study and the design analyses performed during the development of C-E's System 80 plant indicate that the use of suitably designed plutonium burners at Nuclear Energy Centers is technically feasible.

  19. PWR fuel in Japan; Progress and future trends

    Energy Technology Data Exchange (ETDEWEB)

    Yokote, Mitsuhiro (Kansai Electric Power Co., Inc., Osaka (Japan)); Kondo, Yoshiaki; Abeta, Sadaaki (Mitsubishi Heavy Industries Ltd., Tokyo (Japan))

    1994-06-01

    Twenty years ago, in the early years of the Japanese civil nuclear power programme, the fuel used was imported from Westinghouse in the USA. However, it was always intended that there would be a move towards fuel fabrication in Japan and by the end of 1993 around 10,000 Mitsubishi PWR fuel assemblies had been supplied to 21 PWRs in Japan. The highest burnup achieved so far is 46 GWd/t. Design changes to reduce abnormalities have been made, reliability is improving all the time and further improvements in burnup are being developed. This progress in PWR cores and fuel including MOX fuel in Japan is charted and future research and development is outlined. (UK).

  20. A concept of PWR using plate and shell heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Freire, Luciano Ondir; Andrade, Delvonei Alves de, E-mail: luciano.ondir@gmail.com, E-mail: delvonei@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    In previous work it was verified the physical possibility of using plate and shell heat exchangers for steam generation in a PWR for merchant ships. This work studies the possibility of using GESMEX commercial of the shelf plate and shell heat exchanger of series XPS. It was found it is feasible for this type of heat exchanger to meet operational and accidental requirements for steam generation in PWR. Additionally, it is proposed an arrangement of such heat exchangers inside the reactor pressure vessel. Such arrangement may avoid ANSI/ANS51.1 nuclear class I requirements on those heat exchangers because they are contained in the reactor coolant pressure barrier and play no role in accidental scenarios. Additionally, those plates work under compression, preventing the risk of rupture. Being considered non-nuclear safety, having a modular architecture and working under compression may turn such architectural choice a must to meet safety objectives with improved economics. (author)

  1. Control of corrosion product transport in PWR secondary cycles

    Energy Technology Data Exchange (ETDEWEB)

    Sawochka, S.G.; Pearl, W.L. [NWT Corp., San Josa, CA (United States); Passell, T.O.; Welty, C.S. [Electric Power Research Institute, Palo Alto, CA (United States)

    1992-12-31

    Transport of corrosion products to PWR steam generators by the feedwater leads to sludge buildup on the tubesheets and fouling of tube-to-tube support crevices. In these regions, chemical impurities concentrate and accelerate tubing corrosion. Deposit buildup on the tubes also can lead to power generation limitations and necessitate chemical cleaning. Extensive corrosion product transport data for PWR secondary cycles has been developed employing integrating sampling techniques which facilitate identification of major corrosion product sources and assessments of the effectiveness of various control options. Plant data currently are available for assessing the impact of factors such as pH, pH control additive, materials of construction, blowdown, condensate treatment, and high temperature drains and feedwater filtration.

  2. Stochastic Cooling

    Energy Technology Data Exchange (ETDEWEB)

    Blaskiewicz, M.

    2011-01-01

    Stochastic Cooling was invented by Simon van der Meer and was demonstrated at the CERN ISR and ICE (Initial Cooling Experiment). Operational systems were developed at Fermilab and CERN. A complete theory of cooling of unbunched beams was developed, and was applied at CERN and Fermilab. Several new and existing rings employ coasting beam cooling. Bunched beam cooling was demonstrated in ICE and has been observed in several rings designed for coasting beam cooling. High energy bunched beams have proven more difficult. Signal suppression was achieved in the Tevatron, though operational cooling was not pursued at Fermilab. Longitudinal cooling was achieved in the RHIC collider. More recently a vertical cooling system in RHIC cooled both transverse dimensions via betatron coupling.

  3. Decommissioning of the BR3 PWR

    Energy Technology Data Exchange (ETDEWEB)

    Massaut, V.; Klein, M

    1998-07-01

    The objectives, programme and main achievements of SCK-CEN's decommissioning programme in 1997 are summarised. Particular emphasis is on the BR3 decommissioning project. In 1997, auxiliary equipment and loops were dismantled; concrete antimissile slabs were decontaminated; the radiology of the primary loop was modelled; the quality assurance procedure for dismantling loops and equipment were implemented; a method for the dismantling of the reactor pressure vessel was selected; and contaminated thermal insulation of the primary loop containing asbestos was removed.

  4. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pilgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

    1991-12-01

    Benchmark results of the Dutch PINK working group on the PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs.; 9 figs.; 30 tabs.

  5. Study on thermal-hydraulics during a PWR reflood phase

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tadashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    In-core thermal-hydraulics during a PWR reflood phase following a large-break LOCA are quite unique in comparison with two-phase flow which has been studied widely in previous researches, because the geometry of the flow path is complicated (bundle geometry) and water is at extremely low superficial velocity and almost under stagnant condition. Hence, some phenomena realized during a PWR reflood phase are not understood enough and appropriate analytical models have not been developed, although they are important in a viewpoint of reactor safety evaluation. Therefore, author investigated some phenomena specified as important issues for quantitative prediction, i.e. (1) void fraction in a bundle during a PWR reflood phase, (2) effect of radial core power profile on reflood behavior, (3) effect of combined emergency core coolant injection on reflood behavior, and (4) the core separation into two thermal-hydraulically different regions and the in-core flow circulation behavior observed during a combined injection PWR reflood phase. Further, author made analytical models for these specified issues, and succeeded to predict reflood behaviors at representative types of PWRs, i.e.cold leg injection PWRs and Combined injection PWRs, in good accuracy. Above results were incorporated into REFLA code which is developed at JAERI, and they improved accuracy in prediction and enlarged applicability of the code. In the present study, models were intended to be utilized in a practical use, and hence these models are simplified ones. However, physical understanding on the specified issues in the present study is basic and principal for reflood behavior, and then it is considered to be used in a future advanced code development and improvement. (author). 110 refs.

  6. Chemical System Decontamination at PWR Power Stations Biblis A and B by Advanced System Decontamination by Oxidizing Chemistry (ASDOC-D) Process Technology - 13081

    Energy Technology Data Exchange (ETDEWEB)

    Loeb, Andreas; Runge, Hartmut; Stanke, Dieter [NIS Ingenieurgesellschaft mbH, Industriestrasse 13, 63755 Alzenau (Germany); Bertholdt, Horst-Otto [NCT Consulting, Leonhardstrasse 16-18, 90443 Nuernberg (Germany); Adams, Andreas; Impertro, Michael; Roesch, Josef [RWE Power, 68643 Biblis (Germany)

    2013-07-01

    For chemical decontamination of PWR primary systems the so called ASDOC-D process has been developed and qualified at the German PWR power station Biblis. In comparison to other chemical decontamination processes ASDOC-D offers a number of advantages: - ASDOC-D does not require separate process equipment but is completely operated and controlled by the nuclear site installations. Feeding of chemical concentrates into the primary system is done by means of the site's dosing systems. Process control is performed by standard site instrumentation and analytics. - ASDOC-D safely prevents any formation and precipitation of insoluble constituents - Since ASDOC-D is operated without external equipment there is no need for installation of such equipment in high radioactive radiation surrounding. The radioactive exposure rate during process implementation and process performance may therefore be neglected in comparison to other chemical decontamination processes. - ASDOC-D does not require auxiliary hose connections which usually bear high leakage risk. The above mentioned technical advantages of ASDOC-D together with its cost-effectiveness gave rise to Biblis Power station to agree on testing ASDOC-D at the volume control system of PWR Biblis unit A. By involving the licensing authorities as well as expert examiners into this test ASDOC-D received the official qualification for primary system decontamination in German PWR. As a main outcome of the achieved results NIS received contracts for full primary system decontamination of both units Biblis A and B (each 1.200 MW) by end of 2012. (authors)

  7. Technical basis for the initiation and cessation of environmentally-assisted cracking of low-alloy steels in elevated temperature PWR environments

    Energy Technology Data Exchange (ETDEWEB)

    James, L.A.

    1997-10-01

    The Section 11 Working Group on Flaw Evaluation of the ASME B and PV Code Committee is considering a Code Case to allow the determination of the conditions under which environmentally-assisted cracking of low-alloy steels could occur in PWR primary environments. This paper provides the technical support basis for such an EAC Initiation and Cessation Criterion by reviewing the theoretical and experimental information in support of the proposed Code Case.

  8. Assessment of PWR fuel degradation by post-irradiation examinations and modeling in DEGRAD-1 code; Avaliacao da degradacao de combustivel PWR por exames pos-irradiacao e modelagem no codigo DEGRAD-1

    Energy Technology Data Exchange (ETDEWEB)

    Castanheira, Myrthes; Lucki, Georgi; Silva, Jose Eduardo Rosa da; Terremoto, Luis A.A.; Silva, Antonio Teixeira e; Teodoro, Celso A.; Damy, Margaret de A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear]. E-mail: myrthes@ipen

    2005-07-01

    On the majority of the cases, the inquiries on primary failures and secondary in PWR fuel rods are based on results of analysis were made use of the non-destructive examination results (coolant activities monitoring, sipping tests, visual examination). The complementary analysis methodology proposed in this work includes a modeling approach to characterization of the physical effects of the individual chemistry mechanisms that constitute the incubation phase of degradation phenomenon after primary failure that are integrated in the reactor operational history under stationary operational regime, and normal power transients. The computational program called DEGRAD-1 was developed based on this modeling approach. The practical outcome of the program is to predict cladding regions susceptible to massive hydriding. The applications presented demonstrate the validity of proposed method and models by actual cases simulation, which (primary and secondary) defects positions were known and formation time was estimated. By using the modeling approach, a relationship between the hydrogen concentration in the gap and the inner cladding oxide thickness has been identified which, when satisfied, will induce massive hydriding. The novelty in this work is the integrated methodology, which supplements the traditional analysis methods (using data from non-destructive techniques) with mathematical models for the hydrogen evolution, oxidation and hydriding that include refined approaches and criteria for PWR fuel, and using the FRAPCON-3 fuel performance code as the basic tool. (author)

  9. FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

    Directory of Open Access Journals (Sweden)

    Thiollay Nicolas

    2016-01-01

    Full Text Available FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10−2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006–2007 in a geometry representative of 1300 MWe PWR.

  10. PWR Cross Section Libraries for ORIGEN-ARP

    Energy Technology Data Exchange (ETDEWEB)

    McGraw, Carolyn [Texas A& M University; Ilas, Germina [ORNL

    2012-01-01

    New pressurized water reactor (PWR) cross-section libraries were generated for use with the ORIGEN-ARP depletion sequence in the SCALE nuclear analysis code system. These libraries are based on ENDF/B-VII nuclear data and were generated using the two-dimensional depletion sequence, TRITON/NEWT, in SCALE 6.1. The libraries contain multiple burnup-dependent cross-sections for seven PWR fuel designs, with enrichments ranging from 1.5 to 6 wt% 235U. The burnup range has been extended from the 72 GWd/MTU used in previous versions of the libraries to 90 GWd/MTU. Validation of the libraries using radiochemical assay measurements and decay heat measurements for PWR spent fuel showed good agreement between calculated and experimental data. Verification against detailed TRITON simulations for the considered assembly designs showed that depletion calculations performed in ORIGEN-ARP with the pre-generated libraries provide similar results as obtained with direct TRITON depletion, while greatly reducing the computation time.

  11. FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

    Science.gov (United States)

    Thiollay, Nicolas; Di Salvo, Jacques; Sandrin, Charlotte; Soldevila, Michel; Bourganel, Stéphane; Fausser, Clément; Destouches, Christophe; Blaise, Patrick; Domergue, Christophe; Philibert, Hervé; Bonora, Jonathan; Gruel, Adrien; Geslot, Benoit; Lamirand, Vincent; Pepino, Alexandra; Roche, Alain; Méplan, Olivier; Ramdhane, Mourad

    2016-02-01

    FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10-2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006-2007 in a geometry representative of 1300 MWe PWR.

  12. Validation of gadolinium burnout using PWR benchmark specification

    Energy Technology Data Exchange (ETDEWEB)

    Oettingen, Mikołaj, E-mail: moettin@agh.edu.pl; Cetnar, Jerzy, E-mail: cetnar@mail.ftj.agh.edu.pl

    2014-07-01

    Graphical abstract: - Highlights: • We present methodology for validation of gadolinium burnout in PWR. • We model 17 × 17 PWR fuel assembly using MCB code. • We demonstrate C/E ratios of measured and calculated concentrations of Gd isotopes. • The C/E for Gd154, Gd156, Gd157, Gd158 and Gd160 shows good agreement of ±10%. • The C/E for Gd152 and Gd155 shows poor agreement below ±10%. - Abstract: The paper presents comparative analysis of measured and calculated concentrations of gadolinium isotopes in spent nuclear fuel from the Japanese Ohi-2 PWR. The irradiation of the 17 × 17 fuel assembly containing pure uranium and gadolinia bearing fuel pins was numerically reconstructed using the Monte Carlo Continuous Energy Burnup Code – MCB. The reference concentrations of gadolinium isotopes were measured in early 1990s at Japan Atomic Energy Research Institute. It seems that the measured concentrations were never used for validation of gadolinium burnout. In our study we fill this gap and assess quality of both: applied numerical methodology and experimental data. Additionally we show time evolutions of infinite neutron multiplication factor K{sub inf}, FIMA burnup, U235 and Gd155–Gd158. Gadolinium-based materials are commonly used in thermal reactors as burnable absorbers due to large neutron absorption cross-section of Gd155 and Gd157.

  13. PWR core stablity aganst xenon-induced spatial power oscillation

    Energy Technology Data Exchange (ETDEWEB)

    Moon, H.J.; Han, K.I. (Korea Advanced Energy Research Inst., Seoul (Republic of Korea))

    1982-06-01

    Stability of a PWR core against xenon-induced axial power oscillation is studied using one-dimensional xenon transient analysis code, DD1D, that has been developed and verified at KAERI. Analyzed by DD1D utilizing the Kori Unit 1 design and operating data is the sensitivity of axial stability in a PWR core to the changes in core physical parameters including core power level, moderator temperature coefficient, core inlet temperature, doppler power coefficient and core average burnup. Through the sensitivity study the Kori Unit 1 core is found to be stable against axial xenon oscillation at the beginning of cycle 1. But, it becomes less stable as burnup progresses, and unstable at the end of cycle. Such a decrease in stability is mainly due to combined effect of changes in axial power distribution, moderator temperature coefficient and doppler power coefficient as core burnup progresses. It is concluded from the stability analysis of the Kori Unit 1 core that design of a large PWR with high power density and increased dimension can not avoid xenon-induced axial power instabilites to some extents, especially at the end of cycle.

  14. Actinides transmutation - a comparison of results for PWR benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Claro, Luiz H. [Instituto de Estudos Avancados (IEAv/CTA), Sao Jose dos Campos, SP (Brazil)], e-mail: luizhenu@ieav.cta.br

    2009-07-01

    The physical aspects involved in the Partitioning and Transmutation (P and T) of minor actinides (MA) and fission products (FP) generated by reactors PWR are of great interest in the nuclear industry. Besides these the reduction in the storage of radioactive wastes are related with the acceptability of the nuclear electric power. From the several concepts for partitioning and transmutation suggested in literature, one of them involves PWR reactors to burn the fuel containing plutonium and minor actinides reprocessed of UO{sub 2} used in previous stages. In this work are presented the results of the calculations of a benchmark in P and T carried with WIMSD5B program using its new cross sections library generated from the ENDF-B-VII and the comparison with the results published in literature by other calculations. For comparison, was used the benchmark transmutation concept based in a typical PWR cell and the analyzed results were the k{infinity} and the atomic density of the isotopes Np-239, Pu-241, Pu-242 and Am-242m, as function of burnup considering discharge of 50 GWd/tHM. (author)

  15. Recommendations of the MRP-139: Inspection of Welds dissimilar in Nozzles PWR reactor vessel in Spain; Recomendaciones del MRP-139: Inspeccion de soldaduras disimilares en Vasijas de Reactor en Espana

    Energy Technology Data Exchange (ETDEWEB)

    Gadea, J. R.; Willke, A.; Regidor, J. J.; Tecnatom, S. A.

    2010-07-01

    The guide EPRI MRP-139, which provides the way forward for the inspection and evaluation of dissimilar butt welds, the primary system of PWR reactors, indicating the type of nondestructive testing to be done in these areas, based on discovered several cases of default in lnconel alloys 600 and 182 in American and European plants. The phenomenon of cracking.

  16. Analysis of SBO ATWS for Maanshan PWR

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Che-Hao; Chen, Shao-Wen [National Tsing Hua Univ., Hsinchu, Taiwan (China). Inst. of Nuclear Engineering and Science; Wang, Jong-Rong; Shih, Chunkuan [National Tsing Hua Univ., Hsinchu, Taiwan (China). Inst. of Nuclear Engineering and Science; Nuclear and New Energy Education and Research Foundation, Hsinchu, Taiwan (China); Lin, Hao-Tzu [Atomic Energy Council, Taoyuan, Taiwan (China). Inst. of Nuclear Energy Research

    2015-11-15

    Station blackout anticipated transient without scram (SBO ATWS) is considered as loss of off-site and on-site power but no credit for automatic reactor trip. SBO ATWS causes reactor coolant pump (RCP) trip, loss of all main feedwater pumps and turbine trip, then the reactor coolant system (RCS) pressure rises rapidly due to loss of heat removal paths. The ASME Code Level C service limit criteria of 22.06 MPa (3200 psig) is assumed to be an unacceptable plant condition in SECY-83-293. The simulation is performed by TRACE which is a thermal-hydraulic code developed by U.S. NRC. Three different AFW flows are modeled to ensure the pressures will not be beyond the criteria. RCP seal-leakage is concerned as a SBLOCA due to loss of RCP seal-cooling. Four possible leakage flows are modeled to examine the reactor core water level and temperature variation.

  17. PSA LEVEL 3 DAN IMPLEMENTASINYA PADA KAJIAN KESELAMATAN PWR

    Directory of Open Access Journals (Sweden)

    Pande Made Udiyani

    2015-03-01

    Full Text Available Kajian keselamatan PLTN menggunakan metodologi kajian probabilistik sangat penting selain kajian deterministik. Metodologi kajian menggunakan Probabilistic Safety Assessment (PSA Level 3 diperlukan terutama untuk estimasi kecelakaan parah atau kecelakaan luar dasar desain PLTN. Metode ini banyak dilakukan setelah kejadian kecelakaan Fukushima. Dalam penelitian ini dilakukan implementasi PSA Level 3 pada kajian keselamatan PWR, postulasi kecelakan luar dasar desain PWR AP-1000 dan disimulasikan di contoh tapak Bangka Barat. Rangkaian perhitungan yang dilakukan adalah: menghitung suku sumber dari kegagalan teras yang terjadi, pemodelan kondisi meteorologi tapak dan lingkungan, pemodelan jalur paparan, analisis dispersi radionuklida dan transportasi fenomena di lingkungan, analisis deposisi radionuklida, analisis dosis radiasi, analisis perlindungan & mitigasi, dan analisis risiko. Kajian menggunakan rangkaian subsistem pada perangkat lunak PC Cosyma. Hasil penelitian membuktikan bahwa implementasi metode kajian keselamatan PSA Level 3 sangat efektif dan komprehensif terhadap estimasi dampak, konsekuensi, risiko, kesiapsiagaan kedaruratan nuklir (nuclear emergency preparedness, dan manajemen kecelakaan reaktor terutama untuk kecelakaan parah atau kecelakaan luar dasar desain PLTN. Hasil kajian dapat digunakan sebagai umpan balik untuk kajian keselamatan PSA Level 1 dan PSA Level 2. Kata kunci: PSA level 3, kecelakaan, PWR   Reactor safety assessment of nuclear power plants using probabilistic assessment methodology is most important in addition to the deterministic assessment. The methodology of Level 3 Probabilistic Safety Assessment (PSA is especially required to estimate severe accident or beyond design basis accidents of nuclear power plants. This method is carried out after the Fukushima accident. In this research, the postulations beyond design basis accidentsof PWR AP - 1000 would be taken, and simulated at West Bangka sample site. The

  18. PWR FLECHT SEASET 163-Rod Bundle Flow Blockage Task data report. NRC/EPRI/Westinghouse report No. 13, August-October 1982

    Energy Technology Data Exchange (ETDEWEB)

    Loftus, M J; Hochreiter, L E; McGuire, M F; Valkovic, M M

    1983-10-01

    This report presents data from the 163-Rod Bundle Blow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Systems Effects and Separate Effects Test Program (FLECHT SEASET). The task consisted of forced and gravity reflooding tests utilizing electrical heater rods with a cosine axial power profile to simulate PWR nuclear core fuel rod arrays. These tests were designed to determine effects of flow blockage and flow bypass on reflooding behavior and to aid in the assessment of computational models in predicting the reflooding behavior of flow blockage in rod bundle arrays.

  19. IPSN expert appraisal programme on the chooz A 300 MWe PWR. Lessons learned by IPSN

    Energy Technology Data Exchange (ETDEWEB)

    Morlent, O.; Reuchet, J. [CEA Fontenay-aux-Roses, Inst. de Protection et de Surete Nucleaire, 92 (France)

    2001-07-01

    The closure of Chooz A PWR provided an opportunity to take samples of items that had aged in situ in conditions close to those encountered in PWR in operation over a period of 140.000 hours, which is far longer than the usual time-spans of simulated laboratory tests. 4 topics have been studied: 1) effect of radiation on reactor vessel internals, 2) dissimilar metal joints of reactor coolant system: pressurizer surge line, 3) cast parts of austeno-ferritic steel: hot and cold leg primary valves, and 4) ageing of cables in high temperatures and under irradiation. The examination of the lower internals on some baffle angle bracket and core shroud screws, subjected to varying amounts of irradiation, did not reveal any cracking or corrosion, and confirmed the saturation effect between 4 and 10 dpa for the hardening of 304 austenitic steel in the low temperature range. Expert appraisal of the dissimilar metal joints on the pressurizer surge line confirmed the existence of small fabrication defects due to high temperature cracking. Expert appraisal of the 3 valve body samples from the main section of the coolant system confirmed that -) thermal ageing of the valve body on the hot leg was more advanced than that of the cold leg valve, -) the material of the valve housing on the cold leg which, in theory, was not sensitive to ageing phenomena, exhibited unexpectedly low impact strength values. As for cables, measurements confirmed that their mechanical and electrical properties remained sufficient for them to carry out their functions. (A.C.)

  20. EPRI PWR Safety and Relief Valve Test Program: test condition justification report

    Energy Technology Data Exchange (ETDEWEB)

    Hosler, J.

    1982-12-01

    In response to NUREG 0737, Item II.D.1.A requirements, several safety and relief valve designs were tested by EPRI under PWR utility sponsorship. Justification that the inlet fluid conditions under which these valve designs were tested are representative of those expected in participating domestic PWR units during FSAR, Extended High Pressure Injection, and Cold Overpressurization events is presented.

  1. Safety and licensing issues that are being addressed by the Power Burst Facility test programs. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    McCardell, R.K.; MacDonald, P.E.

    1980-01-01

    This paper presents an overview of the results of the experimental program being conducted in the Power Burst Facility and the relationship of these results to certain safety and licensing issues. The safety issues that were addressed by the Power-Cooling-Mismatch, Reactivity Initiated Accident, and Loss of Coolant Accident tests, which comprised the original test program in the Power Burst Facility, are discussed. The resolution of these safety issues based on the results of the thirty-six tests performed to date, is presented. The future resolution of safety issues identified in the new Power Burst Facility test program which consists of tests which simulate BWR and PWR operational transients, anticipated transients without scram, and severe fuel damage accidents, is described.

  2. Timing analysis of PWR fuel pin failures

    Energy Technology Data Exchange (ETDEWEB)

    Jones, K.R.; Wade, N.L.; Katsma, K.R.; Siefken, L.J. (EG and G Idaho, Inc., Idaho Falls, ID (United States)); Straka, M. (Halliburton NUS, Idaho Falls, ID (United States))

    1992-09-01

    Research has been conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B W) design (Oconee) and a Westinghouse (W) four-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin bumup, axial peaking factor, break size, emergency core cooling system availability, and main coolant pump trip on these times. The analysis was performed using the following codes: FRAPCON-2, for the calculation of steady-state fuel behavior; SCDAP/RELAP5/MOD3 and TRACPF1/MOD1, for the calculation of the transient thermal-hydraulic conditions in the reactor system; and FRAP-T6, for the calculation of transient fuel behavior. In addition to the calculation of fuel pin failure timing, this analysis provides a comparison of the predicted results of SCDAP/RELAP5/MOD3 and TRAC-PFL/MOD1 for large-break LOCA analysis. Using SCDAP/RELAP5/MOD3 thermal-hydraulic data, the shortest time intervals calculated between initiation of containment isolation and fuel pin failure are 10.4 seconds and 19.1 seconds for the B W and W plants, respectively. Using data generated by TRAC-PF1/MOD1, the shortest intervals are 10.3 seconds and 29.1 seconds for the B W and W plants, respectively. These intervals are for a double-ended, offset-shear, cold leg break, using the technical specification maximum peaking factor and applied to fuel with maximum design bumup. Using peaking factors commensurate widi actual bumups would result in longer intervals for both reactor designs. This document also contains appendices A through J of this report.

  3. Hot and Cool Spots of Primary Production, Respiration and 15N Nitrate and Ammonium Uptake: Spatial Heterogeneity in Tropical Streams and Rivers

    Science.gov (United States)

    Dodds, W. K.; Tromboni, F.; Neres-Lima, V.; Zandoná, E.; Moulton, T. P.

    2016-12-01

    While whole-stream measures of metabolism and uptake have become common methods to characterize biogeochemical transport and processing, less is known about how nitrogen (N) uptake, gross primary production (GPP) and ecosystem respiration (ER) covary among different stream substrata as smaller scales. We measured 15N ammonium and nitrate uptake seperately, and GPP and ER of ecosystem compartments (leaves, epilithon, sand-associated biota and macrophytes) in closed circulating chambers in three streams/ rivers of varied size. The streams drain pristine Brazilian Atlantic Rainforest watersheds and are all within a few km of eachother. The smallest stream had dense forest canopy cover; the largest river was almost completely open. GPP could not be detected in the closed canopy stream. Epilithon (biofilms on rocks) was a dominant compartment for GPP and N uptake in the two open streams, and macrophytes rivaled epilithon GPP and N uptake rates in the most open stream. Even though leaves covered only 1-3% of the stream bottom, they could account for around half of all the ER in the streams but almost no N uptake. Sand had minimal rates of N uptake, GPP and R associated with it in all streams due to relatively low organic material content. The data suggest that N uptake, GPP and ER of different substrata are not closely linked over relatively small spatial (dm) scales, and that different biogeochemical processes may map to different hot and cool spots for ecosystem rates.

  4. PWR safety and relief valve test program. Valve selection/juftification report. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1982-12-01

    NUREG 0578 required that full-scale testing be performed on pressurizer safety valves and relief valves representative of those in use or planned for use in PWR plants. To obtain valve performance data for the entire population of PWR plant valves, nine safety valves and ten relief valves were selected as a fully representative set of test valves. Justification that the selected valves represent all PWR plant valves was provided by each safety and relief valve manufacturer. Both the valve selection and justification work was performed as part of the PWR Safety and Relief Valve Test Program conducted by EPRI on behalf of the PWR utilities in response to the recommendations of NUREG 0578 and the requirements of the NRC. Results of the Safety and Relief Valve Selection and Justification effort is documented in this report.

  5. Study of the distribution of hydrogen in a PWR containment with CFD codes; Estudio de la distribucion de hidrogeno en una contencion PWR con codigos CFD

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, G.; Matias, R.; Fernandez, K.; Justo, D.; Bocanegra, R.; Mena, L.; Queral, C.

    2015-07-01

    During a severe accident in a PWR, the hydrogen generated may be distributed in the containment atmosphere and reach the combustion conditions that can cause the containment failure. In this research project, a preliminary study has been done about the capacities of ANSYS Fluent 15.0 and GOTHIC 8.0 to tri dimensional distribution of the hydrogen in a PWR containment during a severe accident. (Author)

  6. Assessment of Severe Accident Depressurization Valve Activation Strategy for Chinese Improved 1000 MWe PWR

    Directory of Open Access Journals (Sweden)

    Ge Shao

    2013-01-01

    Full Text Available To prevent HPME and DCH, SADV is proposed to be added to the pressurizer for Chinese improved 1000 MWe PWR NPP with the reference of EPR design. Rapid depressurization capability is assessed using the mechanical analytical code. Three typical severe accident sequences of TMLB’, SBLOCA, and LOFW are selected. It shows that with activation of the SADV the RCS pressure is low enough to prevent HPME and DCH. Natural circulation at upper RPV and hot leg is considered for the rapid depressurization capacity analysis. The result shows that natural circulation phenomenon results in heat transfer from the core to the pipes in RCS which may cause the creep rupture of pipes in RCS and delays the severe accident progression. Different SADV valve areas are investigated to the influence of depressurization of RCS. Analysis shows that the introduction of SADV with right valve area will delay progression of core degradation to RPV failure. Valve area is to be optimized since smaller SADV area will reduce its effect and too large valve area will lead to excessive loss of water inventory in RCS and makes core degradation progression to RPV failure faster without additional core cooling water sources.

  7. Irradiation Effects Test Series: Test IE-2. Test results report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Allison, C. M.; Croucher, D. W.; Ploger, S. A.; Mehner, A. S.

    1977-08-01

    The report describes the results of a test using four 0.97-m long PWR-type fuel rods with differences in diametral gap and cladding irradiation. The objective of this test was to provide information about the effects of these differences on fuel rod behavior during quasi-equilibrium and film boiling operation. The fuel rods were subjected to a series of preconditioning power cycles of less than 30 kW/m. Rod powers were then increased to 68 kW/m at a coolant mass flux of 4900 kg/s-m/sup 2/. After one hour at 68 kW/m, a power-cooling-mismatch sequence was initiated by a flow reduction at constant power. At a flow of 2550 kg/s-m/sup 2/, the onset of film boiling occurred on one rod, Rod IE-011. An additional flow reduction to 2245 kg/s-m/sup 2/ caused the onset of film boiling on the remaining three rods. Data are presented on the behavior of fuel rods during quasiequilibrium and during film boiling operation. The effects of initial gap size, cladding irradiation, rod power cycling, a rapid power increase, and sustained film boiling are discussed. These discussions are based on measured test data, preliminary postirradiation examination results, and comparisons of results with FRAP-T3 computer model calculations.

  8. Zinc injection in German PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Stellwag, B. [Framatome ANP GmbH, Erlangen (Germany); Juergensen, M. [Kernkraftwerk Obrigheim GmbH (Germany); Wolter, D. [RWE Power AG, Kraftwerk Biblis (Germany)

    2002-07-01

    Zinc injection for further reduction of radiation fields was introduced at Unit B of Biblis Nuclear Power Station in September 1996 and at Obrigheim Nuclear Power Station in February 1998. Zinc injection is still being implemented today at these plants. This paper gives an overview of the experience acquired with the method, including the annual refueling outages in the year 2001. The main topic addressed by the paper is the evolution of dose rates at the primary system and work-related doses since introduction of the method. Reductions in high dose rate areas have meanwhile achieved values of 40 to 50%. Annual collective doses per man-hour spent in the controlled access area of the plant as well as personal doses for specific activities are also decreasing. (authors)

  9. A study on thimble plug removal for PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Song, Dong Soo; Lee, Chang Sup; Lee, Jae Yong; Jun, Hwang Yong [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The thermal-hydraulic effects of removing the RCC guide thimble plugs are evaluated for 8 Westinghouse type PWR plants in Korea as a part of feasibility study: core outlet loss coefficient, thimble bypass flow, and best estimate flow. It is resulted that the best estimate thimble bypass flow increases about by 2% and the best estimate flow increases approximately by 1.2%. The resulting DNBR penalties can be covered with the current DNBR margin. Accident analyses are also investigated that the dropped rod transient is shown to be limiting and relatively sensitive to bypass flow variation. 8 refs., 5 tabs. (Author)

  10. Vertical Drop Of 21-Pwr Waste Package On Unyielding Surface

    Energy Technology Data Exchange (ETDEWEB)

    S. Mastilovic; A. Scheider; S.M. Bennett

    2001-01-29

    The objective of this calculation is to determine the structural response of a 21-PWR (pressurized-water reactor) Waste Package (WP) subjected to the 2-m vertical drop on an unyielding surface at three different temperatures. The scope of this calculation is limited to reporting the calculation results in terms of stress intensities in two different WP components. The information provided by the sketches (Attachment I) is that of the potential design of the type of WP considered in this calculation, and all obtained results are valid for that design only.

  11. New instrumentation of reactor water level for PWR; Nueva Instrumentacion de nivel de agua del reactor para PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kaercher, S.

    2005-07-01

    Today, many PWR reactors are equipped with a reactor water level instrumentation system based on different measurement methods. Due to obsolescence issues, FRAMATOME ANP started to develop and quality a new water level measurement system using heated und unheated thermocouple measurements. the measuring principle is based on the fact that the heat transfer in water is considerably higher than in steam. The electronic cabinet for signal processing is based on a proven technology already developed, qualified and installed by FRAMATOME ANP in several NPPs. It is equipped with and advanced temperature measuring transducer for acquisition and processing of thermocouple signals. (Author)

  12. Life management plants at nuclear power plants PWR; Planes de gestion de vida en centrales nucleares PWR

    Energy Technology Data Exchange (ETDEWEB)

    Esteban, G.

    2014-10-01

    Since in 2009 the CSN published the Safety Instruction IS-22 (1) which established the regulatory framework the Spanish nuclear power plants must meet in regard to Life Management, most of Spanish nuclear plants began a process of convergence of their Life Management Plants to practice 10 CFR 54 (2), which is the current standard of Spanish nuclear industry for Ageing Management, either during the design lifetime of the plant, as well as for Long-Term Operation. This article describe how Life Management Plans are being implemented in Spanish PWR NPP. (Author)

  13. VERA Core Simulator Methodology for PWR Cycle Depletion

    Energy Technology Data Exchange (ETDEWEB)

    Kochunas, Brendan [University of Michigan; Collins, Benjamin S [ORNL; Jabaay, Daniel [University of Michigan; Kim, Kang Seog [ORNL; Graham, Aaron [University of Michigan; Stimpson, Shane [University of Michigan; Wieselquist, William A [ORNL; Clarno, Kevin T [ORNL; Palmtag, Scott [Core Physics, Inc.; Downar, Thomas [University of Michigan; Gehin, Jess C [ORNL

    2015-01-01

    This paper describes the methodology developed and implemented in MPACT for performing high-fidelity pressurized water reactor (PWR) multi-cycle core physics calculations. MPACT is being developed primarily for application within the Consortium for the Advanced Simulation of Light Water Reactors (CASL) as one of the main components of the VERA Core Simulator, the others being COBRA-TF and ORIGEN. The methods summarized in this paper include a methodology for performing resonance self-shielding and computing macroscopic cross sections, 2-D/1-D transport, nuclide depletion, thermal-hydraulic feedback, and other supporting methods. These methods represent a minimal set needed to simulate high-fidelity models of a realistic nuclear reactor. Results demonstrating this are presented from the simulation of a realistic model of the first cycle of Watts Bar Unit 1. The simulation, which approximates the cycle operation, is observed to be within 50 ppm boron (ppmB) reactivity for all simulated points in the cycle and approximately 15 ppmB for a consistent statepoint. The verification and validation of the PWR cycle depletion capability in MPACT is the focus of two companion papers.

  14. PWR fuel performance and burnup extension in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Yokote, M. [Kansai Electric Power Co., Inc., Osaka (Japan); Kondo, Y.; Abeta, S.

    1996-10-01

    Japanese utilities and fuel manufacturers have expanded much of their resources and efforts to maintain a reliable supply of PWR fuel for Japan. In the early 1970s, since the level of knowledge and experience of using fuel was less than now, some problems were encountered. However, their causes were investigated and countermeasures implemented, the design improved and quality control enhanced. The results can already be seen by significantly improved performance of the PWR plants now in operation, frequency of problems was quickly reduced. Since fuel reliability has been improved, the emphasis has shifted to improving economics by increasing burnup and using uranium resources effectively. The maximum discharged burnup was previously limited to 39 GWd/t and STEP1 burnup extension to 48 GWd/t has been gradually developed, while STEP2 burnup extension to 55 GWd/t is started to be demonstrated from 1996. Because resources in Japan are scarce, a policy was selected of conserving and making effective use of these resources by recycling the uranium and plutonium recovered from reactors. Consequently, significant work is being done on the development of MOX fuel and utilization of recovered uranium. (author)

  15. Degradation of fastener in reactor internal of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. W.; Ryu, W. S.; Jang, J. S.; Kim, S. H.; Kim, W. G.; Chung, M. K.; Han, C. H

    2000-03-01

    Main component degraded in reactor internal structure of PWR is fastener such as bolts, stud, cap screw, and pins. The failure of these components may damage nuclear fuel and limits the operation of nuclear reactor. In foreign reactors operated more than 10 years, an increasing number of incidents of degraded thread fasteners have been reported. The degradation of these components impair the integrity of reactor internal structure and limit the life extension of nuclear power plant. To solve the problem of fastener failure, the incidents of failure and main mechanisms should be investigated. the purpose of this state-of-the -art report is to investigate the failure incidents and mechanisms of fastener in foreign and domestic PWR and make a guide to select a proper materials. There is no intent to describe each event in detail in this report. This report covers the failures of fastener and damage mechanisms reported by the licensees of operating nuclear power plants and the applications of plants constructed after 1964. This information is derived from pertinent licensee event report, reportable occurrence reports, operating reactor event memoranda, failure analysis reports, and other relevant documents. (author)

  16. Acceptance test for 900 MWe PWR unit replacement steam generators; Essai de reception des generateurs de vapeur de remplacement des tranches REP 900

    Energy Technology Data Exchange (ETDEWEB)

    Gourguechon, B.

    1993-12-31

    During the first half of 1994, the Gravelines 1 steam generators will be replaced (SG replacement procedure). The new SG`s differ from the former components notably by the alloy used for the tube bundle, in this case, the high chromium content Inconel 690. So, from this standpoint, they are to be considered as PWR 900 replacement SG first models and their thermal efficiency has consequently to be assessed. This will provide an opportunity of ensuring that the performance of the components delivered is in compliance with requirements and of making the necessary provisions if significant deviations are observed. The EFMT branch, which has been in charge of the instrumentation and acceptance of the different SG first models since the first PWR plants were commissioned, will be responsible for the acceptance tests and the ultimate validation of a performance assessment procedure applicable to the future replacement steam generators. The methods and tests proposed for SG expert appraisal are based on consideration of the importance of primary measurement quality for satisfactory SG assessment and of the new test facilities with which the 900 and 1 300 PWR plants are gradually being equipped. These facilities provide an on-site computer environment for tests compatible with the tools (PATTERN, etc.) used at EFMT and in other departments. This test is the first of this kind performed by EFMT and the test facility of a nuclear power plant. (author). 6 figs.

  17. Analysis of a bending test on a full-scale PWR hot leg elbow containing a surface crack

    Energy Technology Data Exchange (ETDEWEB)

    Delliou, P. le [Electricite de France, EDF, 77 - Moret-sur-Loing (France). Dept. MTC; Julisch, P.; Hippelein, K. [Stuttgart Univ. (Germany). Staatliche Materialpruefungsanstalt; Bezdikian, G. [Electricite de France, EDF, 92 - Paris la Defense (France). Direction Production Transport

    1998-11-01

    EDF, in co-operation with Framatome, has conducted a large research programme on the mechanical behaviour of thermally aged cast duplex stainless steel elbows, which are part of the main primary circuit of French PWR. One important task of this programme consisted of testing a full-scale PWR hot leg elbow. The elbow contained a semi-elliptical circumferential notch machined on the outer surface of the intrados as well as casting defects located on the flanks. To simulate the end-of-life condition of the component regarding material toughness, it had undergone a 2400 hours ageing heat treatment at 400 C. The test preparation and execution, as well as the material characterization programme, were committed to MPA. The test was conducted under constant internal pressure and in-plane bending (opening mode) at 200 C. For safety reasons, it took place on an open air-site: the Meppen military test ground. At the maximum applied moment (6000 kN.m), the notch did not initiate. This paper presents the experimental results and the fracture mechanics analysis of the test, based on finite element calculations. (orig.)

  18. Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

    Directory of Open Access Journals (Sweden)

    Andrej Prošek

    2013-01-01

    Full Text Available Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO. Long-term SBO in a pressurized water reactor (PWR leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pumps (RCPs leaks assumed to delay the time before core uncovers and significantly heats up. The most important strategies assumed were primary side depressurization and additional makeup water to reactor coolant system (RCS. For simulations of long term SBO scenarios, including early stages of severe accident sequences, the best estimate RELAP5/MOD3.3 and the verified input model of Krško two-loop PWR were used. The results suggest that for the expected magnitude of RCPs seal leak, the core uncovery during the first seven days could be prevented by using the turbine-driven auxiliary feedwater pump and manually depressurizing the RCS through the secondary side. For larger RCPs seal leaks, in general this is not the case. Nevertheless, the core uncovery can be significantly delayed by increasing RCS depressurization.

  19. Evaluation of fretting failures on PWR fuel by post-irradiation examinations and modeling in the DEGRAD-1 code

    Energy Technology Data Exchange (ETDEWEB)

    Castanheira, Myrthes; Silva, Jose Eduardo Rosa da; Lucki, Georgi; Terremoto, Luis A.A.; Silva, Antonio Teixeira e; Teodoro, Celso A.; Damy, Margaret de A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mail: myrthes@ipen.br

    2007-07-01

    One of the major recognized causes of fuel rod failures is fretting of the clad due to the entrapment of debris in a fuel rod spacer. Such debris, inadvertently dropped into the primary system during maintenance operations, includes various sizes of particles. Intermediate size particles, such as metal cuttings, electrical connectors, metal fittings, pieces of wire, and small nuts and bolts can become trapped between fuel rods in a spacer where hydraulically induced vibrations can cause fretting failure of the fuel rod. An evaluation of debris fretting failure on PWR fuel is presented. The inquiries on fuel rods failures are based on results of analysis using post-irradiation non-destructive examination. The complementary analysis includes a modeling approach by code DEGRAD-1 to characterize the degradation phenomenon after primary failure integrated in the reactor operational history. (author)

  20. HIGH COOLING WATER TEMPERATURE EFFECTS ON DESIGN AND OPERATIONAL SAFETY OF NPPS IN THE GULF REGION

    Directory of Open Access Journals (Sweden)

    BYUNG KOO KIM

    2013-12-01

    Full Text Available The Arabian Gulf region has one of the highest ocean temperatures, reaching above 35 degrees and ambient temperatures over 50 degrees in the summer. Two nuclear power plants (NPP are being introduced in the region for the first time, one at Bushehr (1,000 MWe PWR plant from Russia, and a much larger one at Barakah (4X1,400 MWe PWR from Korea. Both plants take seawater from the Gulf for condenser cooling, having to modify the secondary/tertiary side cooling systems design by increasing the heat transfer surface area from the country of origin. This paper analyses the secondary side of a typical PWR plant operating under the Rankine cycle with a simplified thermal-hydraulic model. Parametric study of ocean cooling temperatures is conducted to estimate thermal efficiency variations and its associated design changes for the secondary side. Operational safety is reviewed to deliver rated power output with acceptable safety margins in line with technical specifications, mainly in the auxiliary systems together with the cooling water temperature. Impact on the Gulf seawater as the ultimate heat sink is considered negligible, affecting only the adjacent water near the NPP site, when compared to the solar radiation on the sea surface.

  1. Containment fan cooler heat transfer calculation during main steam line break for Maanshan PWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw; Kao, Lain-Su, E-mail: lskao@iner.gov.tw

    2013-10-15

    Highlights: • Evaluate component cooling water (CCW) thermal response during MSLB for Maanshan. • Using GOTHIC to calculate CCW temperature and determine time required to boil CCW. • Both convective and condensation heat transfer from the air side are considered. • Boiling will not occur since T{sub B} is sufficiently longer than CCW pump restart time. -- Abstract: A thermal analysis has been performed for the Containment Fan Cooler Unit (FCU) during Main Steam Line Break (MSLB) accident, concurrent with loss of offsite power, for Maanshan PWR plant. The analysis is performed in order to address the waterhammer and two-phase flow issues discussed in USNRC's Generic Letter 96-06 (GL 96-06). Maanshan plant is a twin-unit Westinghouse 3-loop PWR currently operated at rated core thermal power of 2822 MWt for each unit. The design basis for containment temperature is Main Steam Line Break (MSLB) accident at power of 2830.5 MWt, which results in peak vapor temperature of 387.6 °F. The design is such that when MSLB occurs concurrent with loss of offsite power (MSLB/LOOP), both the coolant pump on the secondary side and the fan on the air side of the FCU loose power and coast down. The pump has little inertia and coasts down in 2–3 s, while the FCU fan coasts down over much longer period. Before the pump is restored through emergency diesel generator, there is potential for boiling the coolant in the cooling coils by the high-temperature air/steam mixture entering the FCU. The time to boiling depends on the operating pressure of the coolant before the pump is restored. The prediction of the time to boiling is important because it determines whether there is potential for waterhammer or two-phase flow to occur before the pump is restored. If boiling occurs then there exists steam region in the pipe, which may cause the so called condensation induced waterhammer or column closure waterhammer. In either case, a great amount of effort has to be spent to

  2. Danish Cool

    DEFF Research Database (Denmark)

    Toft, Anne Elisabeth

    2016-01-01

    Danish Cool. Keld Helmer-Petersen, Photography and the Photobook Handout exhibition text in English and Chinese by Anne Elisabeth Toft, Curator The exhibition Danish Cool. Keld Helmer-Petersen, Photography and the Photobook presents the ground-breaking work of late Danish photographer Keld Helmer...

  3. Interface tracking simulations of bubbly flows in PWR relevant geometries

    Energy Technology Data Exchange (ETDEWEB)

    Fang, Jun, E-mail: jfang3@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Rasquin, Michel, E-mail: michel.rasquin@colorado.edu [Aerospace Engineering Department, University of Colorado, Boulder, CO 80309 (United States); Bolotnov, Igor A., E-mail: igor_bolotnov@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States)

    2017-02-15

    Highlights: • Simulations were performed for turbulent bubbly flows in PWR subchannel geometry. • Liquid turbulence is fully resolved by direct numerical simulation approach. • Bubble behavior is captured using level-set interface tracking method. • Time-averaged single- and two-phase turbulent flow statistical quantities are obtained. - Abstract: The advances in high performance computing (HPC) have allowed direct numerical simulation (DNS) approach coupled with interface tracking methods (ITM) to perform high fidelity simulations of turbulent bubbly flows in various complex geometries. In this work, we have chosen the geometry of the pressurized water reactor (PWR) core subchannel to perform a set of interface tracking simulations (ITS) with fully resolved liquid turbulence. The presented research utilizes a massively parallel finite-element based code, PHASTA, for the subchannel geometry simulations of bubbly flow turbulence. The main objective for this research is to demonstrate the ITS capabilities in gaining new insight into bubble/turbulence interactions and assisting the development of improved closure laws for multiphase computational fluid dynamics (M-CFD). Both single- and two-phase turbulent flows were studied within a single PWR subchannel. The analysis of numerical results includes the mean gas and liquid velocity profiles, void fraction distribution and turbulent kinetic energy profiles. Two sets of flow rates and bubble sizes were used in the simulations. The chosen flow rates corresponded to the Reynolds numbers of 29,079 and 80,775 based on channel hydraulic diameter (D{sub h}) and mean velocity. The finite element unstructured grids utilized for these simulations include 53.8 million and 1.11 billion elements, respectively. This has allowed to fully resolve all the turbulence scales and the deformable interfaces of individual bubbles. For the two-phase flow simulations, a 1% bubble volume fraction was used which resulted in 17 bubbles in

  4. Dew Point Evaporative Comfort Cooling

    Science.gov (United States)

    2012-11-01

    220 Figure 140. Water-cooled chilled water plant with primary/secondary...enough to buffer the space by carrying away solar loads in unoccupied volumes, such as ceiling plenums. For rooftop installations, where ceiling...and are significant for the three-month period and generally exceed 68%. Larger chilled water plants with water-cooled condensers can operate with

  5. The preliminary thermal–hydraulic analysis of a water cooled blanket concept design based on RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Guanghuai; Peng, Changhong; Guo, Yun, E-mail: guoyun79@ustc.edu.cn

    2016-11-01

    Highlights: • The superheated steam and PWR schemes are analyzed by RELAP5 code. • The influence of non-uniform heating sources is include. • A supposed slow flow decrease case is discussed and the PWR scheme is better. - Abstract: Water cooled blanket (WCB) is very important in the conceptual design and energy transfer in future fusion power plant. One conceptual design of WCB is under computational testing. RELAP5 code, which is mature and often used in transient analysis in Pressurizer water reactor (PWR), is selected as the simulation tool. The complex inner flow channels and heat sources are simplified according to its thermal–hydraulic characteristics. Then the nodal model for REALP5 is built for approximating the conceptual design. Two typical operating plans, superheated steam scheme and PWR scheme, are analyzed. After some adjustments of the inlet flow resistance coefficients of some flow channels, the reasonable stable conditions of both operation plans can be obtained. The stable fluid and wall temperature distributions and pressure drops are studied. At last, a supposed slow flow decreasing is discussed under two operating conditions separately. According to present results, the superheated steam scheme still needs to be further optimized. The PWR scheme shows a very good safety feature.

  6. Modeling local chemistry in PWR steam generator crevices

    Energy Technology Data Exchange (ETDEWEB)

    Millett, P.J. [EPRI, Palo Alto, CA (United States)

    1997-02-01

    Over the past two decades steam generator corrosion damage has been a major cost impact to PWR owners. Crevices and occluded regions create thermal-hydraulic conditions where aggressive impurities can become highly concentrated, promoting localized corrosion of the tubing and support structure materials. The type of corrosion varies depending on the local conditions, with stress corrosion cracking being the phenomenon of most current concern. A major goal of the EPRI research in this area has been to develop models of the concentration process and resulting crevice chemistry conditions. These models may then be used to predict crevice chemistry based on knowledge of bulk chemistry, thereby allowing the operator to control corrosion damage. Rigorous deterministic models have not yet been developed; however, empirical approaches have shown promise and are reflected in current versions of the industry-developed secondary water chemistry guidelines.

  7. PWR steam generator chemical cleaning, Phase I. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Rothstein, S.

    1978-07-01

    United Nuclear Industries (UNI) entered into a subcontract with Consolidated Edison Company of New York (Con Ed) on August 8, 1977, for the purpose of developing methods to chemically clean the secondary side tube to tube support crevices of the steam generators of Indian Point Nos. 1 and 2 PWR plants. This document represents the first reporting on activities performed for Phase I of this effort. Specifically, this report contains the results of a literature search performed by UNI for the purpose of determining state-of-the-art chemical solvents and methods for decontaminating nuclear reactor steam generators. The results of the search sought to accomplish two objectives: (1) identify solvents beyond those proposed at present by UNI and Con Ed for the test program, and (2) confirm the appropriateness of solvents and methods of decontamination currently in use by UNI.

  8. Identifying thermal cycling mechanisms in PWR branch line piping

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. [EPRI, Charlotte, NC (United States); Keller, J.D.; Bilanin, A.J. [Continuum Dynamics, Inc., Ewing, NJ (United States)

    2002-07-01

    Predicting the onset and the characteristics of thermal cycling in pressurized water reactor (PWR) branch line piping systems is critical to formulation of thermal fatigue screening tools. The complex nature of the underlying thermal-hydraulic phenomena, however, significantly complicates prediction using analytical models or direct numerical simulations. Instead, it is necessary to perform scaled experiments to identify the physical mechanisms and to gather data for formulation of semi-empirical models for the thermal cycling phenomena. Through the EPRI Materials Reliability Program a test program is underway to identify and develop semi-empirical correlations for the physical thermalhydraulic mechanisms that cause thermal cycling in dead-ended PWR branch line piping systems. Three series of tests are being performed in this test program: configuration tests on a representative up-horizontal (UH) branch line piping geometry, configuration tests on a representative down-horizontal (DH) branch line piping geometry, and high Reynolds number tests to assess penetration of secondary flow structures into a dead-ended branch line. Results from UH and DH configuration tests indicate that random turbulence penetration is not sufficient for thermal cycling to occur. Rather a swirling flow structure, representative of a large, 'corkscrew' vortical structure, is required for thermal cycling. Scale tests on the UH configuration have simulated cycling phenomena observed in full-scale plant data and have been used to determine parametric sensitivities in formulating a predictive model for the thermal cycling. Data indicate that the mechanism for thermal cycling in UH configurations is stochastic but scales with the leak rate from the valve. The critical dependent variables are reduced to several non-dimensional scaling curves, resulting in a semiempirical predictive model. This paper discusses the test program and the results obtained to date. Application of these

  9. Characterization of Decommissioned PWR Vessel Internals Material Samples: Tensile and SSRT Testing (Nonproprietary Version)

    Energy Technology Data Exchange (ETDEWEB)

    M.Krug, R.Shogan

    2004-09-01

    Pressurized water reactor (PWR) cores operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs requires detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel (internals) subjected to such conditions. This project studied the effects of reactor service on the mechanical and corrosion properties of samples of baffle plate, former plate, and core barrel from a decommissioned PWR.

  10. Identification and evaluation of PWR in-vessel severe accident management strategies

    Energy Technology Data Exchange (ETDEWEB)

    Dukelow, J S [Pacific Northwest Lab., Richland, WA (United States); Harrison, D G [Jason Associates, Idaho Falls, ID (United States); Morgenstern, M [Battelle Human Affairs Research Center, Seattle, WA (United States)

    1992-03-01

    This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents.

  11. Characterization of Decommissioned PWR Vessel Internals Material Samples: Tensile and SSRT Testing (Nonproprietary Version)

    Energy Technology Data Exchange (ETDEWEB)

    M.Krug, R.Shogan

    2004-09-01

    Pressurized water reactor (PWR) cores operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs requires detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel (internals) subjected to such conditions. This project studied the effects of reactor service on the mechanical and corrosion properties of samples of baffle plate, former plate, and core barrel from a decommissioned PWR.

  12. Severe accident analysis in a two-loop PWR nuclear power plant with the ASTEC code

    Energy Technology Data Exchange (ETDEWEB)

    Sadek, Sinisa; Amizic, Milan; Grgic, Davor [Zagreb Univ. (Croatia). Faculty of Electrical Engineering and Computing

    2013-12-15

    The ASTEC/V2.0 computer code was used to simulate a hypothetical severe accident sequence in the nuclear power plant Krsko, a 2-loop pressurized water reactor (PWR) plant. ASTEC is an integral code jointly developed by Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Germany) to assess nuclear power plant behaviour during a severe accident. The analysis was conducted in 2 steps. First, the steady state calculation was performed in order to confirm the applicability of the plant model and to obtain correct initial conditions for the accident analysis. The second step was the calculation of the station blackout accident with a leakage of the primary coolant through degraded reactor coolant pump seals, which was a small LOCA without makeup capability. Two scenarios were analyzed: one with and one without the auxiliary feedwater (AFW). The latter scenario, without the AFW, resulted in earlier core damage. In both cases, the accident ended with a core melt and a reactor pressure vessel failure with significant release of hydrogen. In addition, results of the ASTEC calculation were compared with results of the RELAP5/SCDAPSIM calculation for the same transient scenario. The results comparison showed a good agreement between predictions of those 2 codes. (orig.)

  13. Regulatory Research of the PWR Severe Accident. Information Needs and Instrumentation for Hydrogen Control and Management

    Energy Technology Data Exchange (ETDEWEB)

    Park, Gun Chul; Suh, Kune Y.; Lee, Jin Yong; Lee, Seung Dong [Seoul Nat' l Univ., Seoul (Korea, Republic of)

    2001-03-15

    The current research is concerned with generation of basic engineering data needed in the process of developing hydrogen control guidelines as part of accident management strategies for domestic nuclear power plants and formulating pertinent regulatory requirements. Major focus is placed on identification of information needs and instrumentation methods for hydrogen control and management in the primary system and in the containment, development of decision-making trees for hydrogen management and their quantification, the instrument availability under severe accident conditions, critical review of relevant hydrogen generation model and phenomena In relation to hydrogen behavior, we analyzed the severe accident related hydrogen generation in the UCN 3{center_dot}4 PWR with modified hydrogen generation model. On the basis of the hydrogen mixing experiment and related GASFLOW calculation, the necessity of 3-dimensional analysis of the hydrogen mixing was investigated. We examined the hydrogen control models related to the PAR(Passive Autocatalytic Recombiner) and performed MAAP4 calculation in relation to the decision tree to estimate the capability and the role of the PAR during a severe accident.

  14. Development of a parametric containment event tree model of a severe PWR accident

    Energy Technology Data Exchange (ETDEWEB)

    Okkonen, T. [OTO-Consulting Ay, Helsinki (Finland)

    1996-06-01

    The study supports the development project of STUK on `Living` PSA Level 2. The main work objective is to develop review tools for the Level 2 PSA studies underway at the utilities. The SPSA (STUK PSA) code is specifically designed for the purpose. In this work, SPSA is utilized as the Level 2 programming and calculation tool. A containment event tree (CET) model is built for analysis of severe accidents at the Loviisa pressurized water reactor (PWR) units. Parametric models of severe accident progression and fission product behaviour are developed and integrated in order to construct a compact and self-contained Level 2 PSA model. The model can be easily updated to include new research results, and so it facilitates the Living PSA concept on Level 2 as well. The analyses of the study are limited to severe accidents starting from full-power operation and leading to core melting at a low primary system pressure. Severe accident progression from five plant damage states (PDSs) is examined, however the integration with Level 1 is deferred to more definitive, integrated, safety assessments. (34 refs., 5 figs., 9 tabs.).

  15. VERA-CS Modeling and Simulation of PWR Main Steam Line Break Core Response to DNB

    Energy Technology Data Exchange (ETDEWEB)

    Salko, Robert K [ORNL; Sung, Yixing [Westinghouse Electric Company, Cranberry Township; Kucukboyaci, Vefa [Westinghouse Electric Company, Cranberry Township; Xu, Yiban [Westinghouse Electric Company, Cranberry Township; Cao, Liping [Westinghouse Electric Company, Cranberry Township

    2016-01-01

    The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time step of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.

  16. PWR neutron ex-vessel detection calculations using three-dimensional codes; Calculs de detection neutronique externe dans un rep

    Energy Technology Data Exchange (ETDEWEB)

    Dekens, O.; Lefebvre, J.C.; Rohart, M. [Electricite de France (EDF), 69 -Villeurbanne (France); Chiron, M. [CEA Centre d`Etudes de Saclay, 91 -Gif-sur-Yvette (France). Direction des Reacteurs Nucleaires; Wouters, R. de [TRACTEBEL, Brussels (Belgium)

    1997-10-01

    During the accident of TM12, the signal delivered by source detectors was exceptionally high. This phenomenon was found out to be due to the water inventory in the primary system. Thus, in their research activity, Electricite de France (EdF) and Commissariat a l`Energie Atomique (CEA) have jointly launched a programme, whose aim was to determine to what extent the response of ex-vessel neutron detectors are representative of reactor water level (or sources positions) in a French 900 MWe PWR. In this framework, both partners developed the methods needed for each step of the calculation chain. Finally, a simulation of a LOCA indicates that the loss of coolant can be detected by existing monitoring system, and could be more efficiently found by changing the position of the source range detectors. (authors). 11 refs.

  17. PWR neutron ex-vessel detection calculations using three-dimensional codes; Calculs de detection neutronique externe dans un rep

    Energy Technology Data Exchange (ETDEWEB)

    Dekens, O.; Lefebvre, J.C.; Rohart, M. [Electricite de France (EDF), 69 -Villeurbanne (France); Chiron, M. [CEA Centre d`Etudes de Saclay, 91 -Gif-sur-Yvette (France). Direction des Reacteurs Nucleaires; Wouters, R. de [TRACTEBEL, Brussels (Belgium)

    1997-10-01

    During the accident of TM12, the signal delivered by source detectors was exceptionally high. This phenomenon was found out to be due to the water inventory in the primary system. Thus, in their research activity, Electricite de France (EdF) and Commissariat a l`Energie Atomique (CEA) have jointly launched a programme, whose aim was to determine to what extent the response of ex-vessel neutron detectors are representative of reactor water level (or sources positions) in a French 900 MWe PWR. In this framework, both partners developed the methods needed for each step of the calculation chain. Finally, a simulation of a LOCA indicates that the loss of coolant can be detected by existing monitoring system, and could be more efficiently found by changing the position of the source range detectors. (authors). 11 refs.

  18. Experiment data report for Semiscale Mod-1 Tests S-28-7, S-28-9, and S-28-12. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Esparza, V.; Collins, B.L.; Sackett, K.E.; Coppin, C.E.

    1978-02-01

    Recorded test data are presented for Tests S-28-7, S-28-9, and S-28-12 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Tests S-28-7, S-28-9, and S-28-12 were conducted from initial conditions of 15 736 kPa and 557 K, 15 754 kPa and 556 K, and 15 704 kPa and 559 K, respectively, to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. The specific objective of these tests was to refine the definition of the upper limit of steam generator tube ruptures at which high peak cladding temperatures occur, as set by Test S-28-1. During these tests, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant in a PWR. Thirty (Test S-28-7), 34 (Test S-28-9), and 20 (Test S-28-12) steam generator tube ruptures were simulated by a controlled injection from a heated accmulator into the intact loop hot leg.

  19. GOTHIC Simulation of Passive Containment Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Huiun; Kim, Hangon [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2013-05-15

    The performance of this system depends on the condensation of steam moving downward inside externally cooled vertical tubes. AES-2006: During a DBA, heat is removed by internally cooled vertical tubes, which are located in containment. We are currently developing the conceptual design of Innovative PWR, which is will be equipped with various passive safety features, including PCCS. We have plan to use internal heat exchanger (HX) type PCCS with concrete containment. In this case, the elevation of HXs is important to ensure the heat removal during accidents. In general, steam is lighter than air mixture in containment. So, steam may be collected at the upper side of containment. It means that higher elevation of HXs, larger heat removal efficiency of those. So, the aim of the present paper is to give preliminary study on variation of heat removal performance according to elevation of HXs. With reference to the design specification of the current reactors including APR+, we had determined conceptual design of PCCS. Using it, we developed a GOTHIC model of the APR1400 containment was adopted PCCS. This calculation model is described herein and representative results of calculation are presented. APR 1400 GOTHIC model was developed for PCCS performance calculation and sensitivity test according to installation elevation of PCCXs. Calculation results confirm that PCCS is working properly. It is found that the difference due to the installation elevation of PCCXs is insignificant at this preliminary analysis, however, further studies should be performed to confirm final performance of PCCS according to the installation elevation. These insights are important for developing the PCCS of Innovative PWR.

  20. Ventilative Cooling

    DEFF Research Database (Denmark)

    Heiselberg, Per Kvols; Kolokotroni, Maria

    This report, by venticool, summarises the outcome of the work of the initial working phase of IEA ECB Annex 62 Ventilative Cooling and is based on the findings in the participating countries. It presents a summary of the first official Annex 62 report that describes the state-of-the-art of ventil......This report, by venticool, summarises the outcome of the work of the initial working phase of IEA ECB Annex 62 Ventilative Cooling and is based on the findings in the participating countries. It presents a summary of the first official Annex 62 report that describes the state...

  1. Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J. L. [Rempe and Associates, LLC, Idaho Falls, ID (United States); Knudson, D. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lutz, R. J. [Lutz Nuclear Safety Consultant, LLC, Asheville, NC (United States)

    2015-09-01

    The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure that critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that

  2. Cool snacks

    DEFF Research Database (Denmark)

    Grunert, Klaus G; Brock, Steen; Brunsø, Karen

    2016-01-01

    such a product requires an interdisciplinary effort where researchers with backgrounds in psychology, anthropology, media science, philosophy, sensory science and food science join forces. We present the COOL SNACKS project, where such a blend of competences was used first to obtain thorough insight into young...

  3. Stochastic cooling

    Energy Technology Data Exchange (ETDEWEB)

    Bisognano, J.; Leemann, C.

    1982-03-01

    Stochastic cooling is the damping of betatron oscillations and momentum spread of a particle beam by a feedback system. In its simplest form, a pickup electrode detects the transverse positions or momenta of particles in a storage ring, and the signal produced is amplified and applied downstream to a kicker. The time delay of the cable and electronics is designed to match the transit time of particles along the arc of the storage ring between the pickup and kicker so that an individual particle receives the amplified version of the signal it produced at the pick-up. If there were only a single particle in the ring, it is obvious that betatron oscillations and momentum offset could be damped. However, in addition to its own signal, a particle receives signals from other beam particles. In the limit of an infinite number of particles, no damping could be achieved; we have Liouville's theorem with constant density of the phase space fluid. For a finite, albeit large number of particles, there remains a residue of the single particle damping which is of practical use in accumulating low phase space density beams of particles such as antiprotons. It was the realization of this fact that led to the invention of stochastic cooling by S. van der Meer in 1968. Since its conception, stochastic cooling has been the subject of much theoretical and experimental work. The earliest experiments were performed at the ISR in 1974, with the subsequent ICE studies firmly establishing the stochastic cooling technique. This work directly led to the design and construction of the Antiproton Accumulator at CERN and the beginnings of p anti p colliding beam physics at the SPS. Experiments in stochastic cooling have been performed at Fermilab in collaboration with LBL, and a design is currently under development for a anti p accumulator for the Tevatron.

  4. Industry Application ECCS / LOCA Integrated Cladding/Emergency Core Cooling System Performance: Demonstration of LOTUS-Baseline Coupled Analysis of the South Texas Plant Model

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Szilard, Ronaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Epiney, Aaron [Idaho National Lab. (INL), Idaho Falls, ID (United States); Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Vaghetto, Rodolfo [Texas A & M Univ., College Station, TX (United States); Vanni, Alessandro [Texas A & M Univ., College Station, TX (United States); Neptune, Kaleb [Texas A & M Univ., College Station, TX (United States)

    2017-06-01

    Under the auspices of the DOE LWRS Program RISMC Industry Application ECCS/LOCA, INL has engaged staff from both South Texas Project (STP) and the Texas A&M University (TAMU) to produce a generic pressurized water reactor (PWR) model including reactor core, clad/fuel design and systems thermal hydraulics based on the South Texas Project (STP) nuclear power plant, a 4-Loop Westinghouse PWR. A RISMC toolkit, named LOCA Toolkit for the U.S. (LOTUS), has been developed for use in this generic PWR plant model to assess safety margins for the proposed NRC 10 CFR 50.46c rule, Emergency Core Cooling System (ECCS) performance during LOCA. This demonstration includes coupled analysis of core design, fuel design, thermalhydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results. Within this context, a multi-physics best estimate plus uncertainty (MPBEPU) methodology framework is proposed.

  5. Mitsubishi PWR nuclear fuel with advanced design features

    Energy Technology Data Exchange (ETDEWEB)

    Kaua Goe, Toshiy Uki; Nuno kawa, Koi Chi [Mitsubishi Heavy Industries, Ltd., Tokyo (Japan)

    2008-10-15

    In the last few decades, the global warming has been a big issue. As the breakthrough in this crisis, advanced operations of the water reactor such as higher burn up, longer cycle, and up rating could be effective ways. From this viewpoint, Mitsubishi Heavy Industries (MHI) has developed the fuel for burn up extension, whose assembly burn-up limit is 55GWd/t(A), with the original and advanced designs such as corrosion resistant cladding material MDA, and supplied to Japanese PWR utilities. On the other hand, MHI intends to supply more advanced fuel assemblies not only to domestic market but to the global market. Actually MHI has submitted the application for standard design certification of USA . Advanced Pressurized Water Reactor on Jan. 2nd 2008. The fuel assembly for US APWR is 17x17 type with active fuel length of 14ft, characterized with three features, to {sup E}nhance Fuel Economy{sup ,} {sup E}nable Flexible Core Operation{sup ,} and to {sup I}mprove Reliability{sup .} MHI has also been conducting development activities for more advanced products, such as 70GWd/t(A) burn up limit fuel with cladding, guide thimble and spacer grid made from M-MDATM alloy that is new material with higher corrosion resistance, such as 12ft and 14ft active length fuel, such as fuel with countermeasure against grid fretting, debris fretting, and IRI. MHI will present its activities and advanced designs.

  6. A very cool cooling system

    CERN Multimedia

    Antonella Del Rosso

    2015-01-01

    The NA62 Gigatracker is a jewel of technology: its sensor, which delivers the time of the crossing particles with a precision of less than 200 picoseconds (better than similar LHC detectors), has a cooling system that might become the precursor to a completely new detector technique.   The 115 metre long vacuum tank of the NA62 experiment. The NA62 Gigatracker (GTK) is composed of a set of three innovative silicon pixel detectors, whose job is to measure the arrival time and the position of the incoming beam particles. Installed in the heart of the NA62 detector, the silicon sensors are cooled down (to about -20 degrees Celsius) by a microfluidic silicon device. “The cooling system is needed to remove the heat produced by the readout chips the silicon sensor is bonded to,” explains Alessandro Mapelli, microsystems engineer working in the Physics department. “For the NA62 Gigatracker we have designed a cooling plate on top of which both the silicon sensor and the...

  7. In-situ oxide layer analysis of alloy 182 using electrochemical impedance spectroscopy in high dissolved hydrogen condition in PWR environment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ho-Sub; Subramanian, Gokul Obulan; Hong, Jong-Dae; Lee, Junho; Jang, Changheui [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    Alloy 82/182 weld metals had been extensively used in joining the components of the PWR primary system. Unfortunately, the cracking caused by PWSCC usually occurs on Alloy 82/182 dissimilar metal welds (DMW). Previous studies indicated that the susceptibility of PWSCC is closely related to the oxide characteristics which are dependent on water chemistry condition, especially dissolved hydrogen (DH). Furthermore, in primary system of pressurized water reactor (PWR), crack initiation resulted from electrochemical instability of oxide film of Ni-base structural materials in various hydrogen concentrations. In this study, in-situ oxide analysis of Alloy 182 using electrochemical impedance spectroscopy (EIS) was performed in high dissolved hydrogen condition. Especially, to understand the effects of tensile loading on the oxide characteristics, we tried to characterize the oxides formed on the tensile loaded specimen using in-situ EIS analysis. The EIS analysis of oxide on Alloy 182 was performed. The increase of oxide film thickness was observed with the increase of exposure time. To analysis the multi-layer structure of oxides, an equivalent model was obtained by fitting EIS data. It is assumed that overall oxide structures were composed of 3 layers approximately.

  8. Cooling technique

    Energy Technology Data Exchange (ETDEWEB)

    Salamon, Todd R; Vyas, Brijesh; Kota, Krishna; Simon, Elina

    2017-01-31

    An apparatus and a method are provided. Use is made of a wick structure configured to receive a liquid and generate vapor in when such wick structure is heated by heat transferred from heat sources to be cooled off. A vapor channel is provided configured to receive the vapor generated and direct said vapor away from the wick structure. In some embodiments, heat conductors are used to transfer the heat from the heat sources to the liquid in the wick structure.

  9. Large Scale Finite Element Thermal Analysis of the Bolts of a French PWR Core Internal Baffle Structure

    Energy Technology Data Exchange (ETDEWEB)

    Rupp, Isabelle; Christophe, Peniguel [EDF R and D, Paris (France); Tommy, Martin Michel [1 av du General de Gaulle, Paris (France)

    2009-11-15

    The internal core baffle structure of a French Pressurized Water Reactor (PWR) consists of a collection of baffles and formers that are attached to the barrel. The connections are done thanks to a large number of bolts (about 1500). After inspection, some of the bolts have been found cracked. This has been attributed to the Irradiation Assisted Stress Corrosion Cracking (IASCC). The Electricite De France (EDF) has set up a research program to gain better knowledge of the temperature distribution, which may affect the bolts and the whole structure. The temperature distribution in the structure was calculated thanks to the thermal code SYRTHES that used a finite element approach. The heat transfer between the by-pass flow inside the cavities of the core baffle and the structure was accounted for thanks to a strong thermal coupling between the thermal code SYRTHES and the CFD code named Code{sub S}aturne. The results for the CP0 plant design show that both the high temperature and strong temperature gradients could potentially induce mechanical stresses. The CPY design, where each bolt is individually cooled, had led to a reduction of temperatures inside the structures. A new parallel version of SYRTHES, for calculations on very large meshes and based on MPI, has been developed. A demonstration test on the complete structure that has led to about 1.1 billion linear tetraedra has been calculated on 2048 processors of the EDF Blue Gene computer

  10. ANALISIS LAJU DOSIS NEUTRON REAKTOR PLTN PWR 1000 MWe MENGGUNAKAN PROGRAM MCNP

    Directory of Open Access Journals (Sweden)

    Amir Hamzah

    2015-03-01

    Full Text Available Dalam rangka menyongsong PLTN pertama di Indonesia, dilakukan kajian dan analisis berbagai aspek teknologi reaktor tersebut. Tujuan dari penelitian ini adalah menentukan laju dosis neutron di luar perisai biologik reaktor PLTN PWR 1000 MWe yang merupakan bagian dari kegiatan besar di atas. Data hasil analisis laju dosis radiasi pada posisi tertentu sangat dibutuhkan untuk menunjukkan tingkat paparan radiasi di posisi tersebut. Analisis laju dosis neutron ditentukan berdasarkan hasil analisis fluks dan spektrum neutron. Analisis fluks dan spektrum neutron di teras reaktor daya PWR 1000 Mwe dilakukan menggunakan program MCNP. Model perhitungan yang dilakukan meliputi 9 zona material yaitu, teras, air, selimut, air, tong, air, bejana tekan, beton dan lapisan udara luar. Penentuan distribusi fluks dan spektrum neutron dilakukan ke arah radial hingga di luar perisai beton dengan akurasi antara 10% hingga 30% dalam tiap kelompok energi yang jumlahnya 1 dan 50 kelompok. Hasil analisis laju dosis neutron di permukaan perisai biologik reaktor PLTN PWR 1000 MWe pada kondisi reaktor beroperasi daya penuh sudah di bawah nilai batas keselamatan. Maka dapat disimpulkan bahwa dari segi paparan radiasi neutron, penggunaan perisai radiasi beton setebal dua meter sudah memenuhi persyaratan keselamatan. Kata kunci: PLTN PWR, fluks neutron, perisai, laju dosis neutron, MCNP.   In order to meet the first nuclear power plant in Indonesia, it has been conducted a study and analysis of various aspects of reactor technology. The purpose of this study was to determine the neutron dose rates at the outside of biological shield of NPP PWR 1000 MWe reactor that is a part of the activities described above. The analysis data of radiation dose rate at a specific position is needed to show the level of radiation exposure in those positions. Analysis neutron dose rate is determined based on the results of the analysis of neutron flux. Analysis of flux and neutron spectrum in

  11. Nonlinear Fuzzy Model Predictive Control for a PWR Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Xiangjie Liu

    2014-01-01

    Full Text Available Reliable power and temperature control in pressurized water reactor (PWR nuclear power plant is necessary to guarantee high efficiency and plant safety. Since the nuclear plants are quite nonlinear, the paper presents nonlinear fuzzy model predictive control (MPC, by incorporating the realistic constraints, to realize the plant optimization. T-S fuzzy modeling on nuclear power plant is utilized to approximate the nonlinear plant, based on which the nonlinear MPC controller is devised via parallel distributed compensation (PDC scheme in order to solve the nonlinear constraint optimization problem. Improved performance compared to the traditional PID controller for a TMI-type PWR is obtained in the simulation.

  12. AREVA solutions to licensing challenges in PWR and BWR reload and safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curca-Tivig, Florin [AREVA GmbH, Erlangen (Germany)

    2016-05-15

    Regulatory requirements for reload and safety analyses are evolving: new safety criteria, request for enlarged qualification databases, statistical applications, uncertainty propagation.. In order to address these challenges and access more predictable licensing processes, AVERA is implementing consistent code and methodology suites for PWR and BWR core design and safety analysis, based on first principles modeling and extremely broad verification and validation data base. Thanks to the high computational power increase in the last decades methods' development and application now include new capabilities. An overview of the main AREVA codes and methods developments is given covering PWR and BWR applications in different licensing environments.

  13. Analyses of PWR boron dilution consequences with the Arrotta code

    Energy Technology Data Exchange (ETDEWEB)

    Johanson, E.; Cheng, H.W.; Sehgal, B.R. [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety

    1998-03-01

    During the past few years, major attention has been paid to analyzing the issue of reactivity initiated accidents (RIAs), of which the boron dilution event is of very special interest to the countries having pressurized water reactors (PWRs) in their nuclear power delivery systems. The scenario considered is that if an inadvertent accumulation of boron free water in one loop during reactor startup operations of a PWR and the inadvertent startup of the reactor coolant pump (RCP) in the loop. This could then lead to a rapid boron dilution in the core, which can in turn give rise to a power excursion. This report is devoted to studying the potential physical and thermal hydraulic consequences of a slug of diluted coolant entering the core after one RCP start under a couple of postulated cases. The severity of the consequences of such a scenario is primarily determined by the amount of positive reactivity insertion, and they are also related to the reactivity insertion rate. Therefore, in the report, detailed calculations and analyses have been carried out from case to case by using the well-known space-time kinetics code, ARROTTA. As a result, the spatial distribution for nodal power, fuel enthalpy, fuel temperature and clad outside temperature as well as the change in core reactivity, total core power and peak fuel temperature can be provided. In general, the maximum fuel enthalpy, peak fuel temperature, and clad outside temperature, for all the cases considered in the report, do not exceed their respective routine safety limitations because of the strong Doppler effect and moderator temperature feedback, except if the safety limitations on fuel enthalpy addition for high burnup fuel are drastically reduced.

  14. Continuous firefly algorithm applied to PWR core pattern enhancement

    Energy Technology Data Exchange (ETDEWEB)

    Poursalehi, N., E-mail: npsalehi@yahoo.com [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Tehran (Iran, Islamic Republic of); Zolfaghari, A.; Minuchehr, A.; Moghaddam, H.K. [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Tehran (Iran, Islamic Republic of)

    2013-05-15

    Highlights: ► Numerical results indicate the reliability of CFA for the nuclear reactor LPO. ► The major advantages of CFA are its light computational cost and fast convergence. ► Our experiments demonstrate the ability of CFA to obtain the near optimal loading pattern. -- Abstract: In this research, the new meta-heuristic optimization strategy, firefly algorithm, is developed for the nuclear reactor loading pattern optimization problem. Two main goals in reactor core fuel management optimization are maximizing the core multiplication factor (K{sub eff}) in order to extract the maximum cycle energy and minimizing the power peaking factor due to safety constraints. In this work, we define a multi-objective fitness function according to above goals for the core fuel arrangement enhancement. In order to evaluate and demonstrate the ability of continuous firefly algorithm (CFA) to find the near optimal loading pattern, we developed CFA nodal expansion code (CFANEC) for the fuel management operation. This code consists of two main modules including CFA optimization program and a developed core analysis code implementing nodal expansion method to calculate with coarse meshes by dimensions of fuel assemblies. At first, CFA is applied for the Foxholes test case with continuous variables in order to validate CFA and then for KWU PWR using a decoding strategy for discrete variables. Results indicate the efficiency and relatively fast convergence of CFA in obtaining near optimal loading pattern with respect to considered fitness function. At last, our experience with the CFA confirms that the CFA is easy to implement and reliable.

  15. French nuclear plants PWR vessel integrity assessment and life management

    Energy Technology Data Exchange (ETDEWEB)

    Bezdikian, G. [Electricite de France (EDF), Div. Production Nucleaire, 93 - Saint-Denis (France); Quinot, P. [FRAMATOME, Dept. Bloc Reacteur et Boucles Primaires, 92 - Paris-La-Defence (France); Faidy, C.; Churier-Bossennec, H. [Electricite de France (EDF), Div. Ingenierie et Service, 69 - Villeurbanne (France)

    2001-07-01

    The Reactor Pressure Vessel life management of 56 PWR 3 loop and 4 loop reactors units was engaged by the French Utility EDF (Electricite de France) a few years ago and is yet on going on. This paper will present the work carried out within the framework of justifying why the 34 three loop reactor vessels will remain acceptable for operation for a lifetime of at least 40-years. A summary of the measures will be given. An overall review of actions will be presented describing the French approach, using important existing databases, including studies related to irradiation surveillance monitoring program and end of life fluence assessment. The last results obtained are based on generic integrity analyses for all categories of situations (normal upset emergency and faulted conditions) until the end of lifetime, postulating circumferential an radial kinds of flaw located in the stainless steel cladding or shallow sub-cladding area. The results of structural integrity analyses beginning with elastic computations and completed with three-dimensional finite element elastic plastic computations for envelope cases, are compared with code criteria for operating plants. The objective is to evaluate the margins on different parameters as RTNDT (Reference Nil Ductility Transition Temperature), toughness or crack size, to justify the global fitness for service of all these Reactor Pressure Vessels. The paper introduces EDF's maintenance strategy, related to integrity assessment, for those nuclear power plants under operation, based on NDE in-service inspection of the first thirty millimeters in the thickness of the wall and major surveillance programs of the vessels. (author)

  16. ATLAS - Liquid Cooling Systems

    CERN Multimedia

    Bonneau, P.

    1998-01-01

    Photo 1 - Cooling Unit - Side View Photo 2 - Cooling Unit - Detail Manifolds Photo 3 - Cooling Unit - Rear View Photo 4 - Cooling Unit - Detail Pump, Heater and Exchanger Photo 5 - Cooling Unit - Detail Pump and Fridge Photo 6 - Cooling Unit - Front View

  17. Topical report on actinide-only burnup credit for PWR spent nuclear fuel packages. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1997-04-01

    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package

  18. Thermal-hydraulic analysis of NSSS and containment response during extended station blackout for Maanshan PWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw; Hsu, Keng-Hsien, E-mail: hardlycampus@iner.gov.tw; Lin, Chin-Tsu, E-mail: jtling@iner.gov.tw

    2015-07-15

    Highlights: • Calculate NSSS and containment transient response during extended SBO of 24 h. • RELAP5-3D and GOTHIC models are developed for Maanshan PWR plant. • Reactor coolant pump seal leakage is specifically modeled for each loop. • Analyses are performed with and without secondary-side depressurization, respectively. • Considering different total available time for turbine driven auxiliary feedwater system. - Abstract: A thermal-hydraulic analysis has been performed with respect to the response of the nuclear steam supply system (NSSS) and the containment during an extended station blackout (SBO) duration of 24 h in Maanshan PWR plant. Maanshan plant is a Westinghouse three-loop PWR design with rated core thermal power of 2822 MWt. The analyses in the NSSS and the containment are based on the RELAP5-3D and GOTHIC models, respectively. Important design features of the plant in response to SBO are considered in the respective models, e.g., the steam generator PORVs, turbine driven auxiliary feedwater system (TDAFWS), accumulators, reactor coolant pump (RCP) seal design, various heat structures in the containment, etc. In the analysis it is assumed that the shaft seal in each RCP failed due to loss of seal cooling and the RCS fluid flows to the containment directly. Some parameters calculated from the RELPA5-3D model are input to the containment GOTHIC model, including the RCS average temperature and the RCP seal leakage flow and enthalpy. The RCS average temperature is used to drive the sensible heat transfer to the containment. It is found that the severity of the event depends mainly on whether the secondary side is depressurized or not. If the secondary side is depressurized in time (within 1 h after SBO) and the TDAFWS is available greater than 19 h, then the reactor core will be covered with water throughout the SBO duration, which ensures the integrity of the reactor core. On the contrary, if the secondary side is not depressurized, then the RCS

  19. Effectiveness of a web-based education program to improve vaccine storage conditions in primary care (Keep Cool): study protocol for a randomized controlled trial

    OpenAIRE

    Thielmann, Anika; Viehmann, Anja; Weltermann, Birgitta M

    2015-01-01

    Background Immunization programs are among the most effective public health strategies worldwide. Adequate vaccine storage is a prerequisite to assure the vaccines’ effectiveness and safety. In a questionnaire survey among a random sample of German primary care physicians, we discovered vaccine storage deficits: 16 % of physicians had experience with cold chain breaches either as an error or near error, 49 % did not keep a temperature log, and 21 % did not use a separate refrigerator for vacc...

  20. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Refeat, Riham Mahmoud [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

    2015-12-15

    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

  1. Criticality safety and sensitivity analyses of PWR spent nuclear fuel repository facilities

    NARCIS (Netherlands)

    Maucec, M; Glumac, B

    2005-01-01

    Monte Carlo criticality safety and sensitivity calculations of pressurized water reactor (PWR) spent nuclear fuel repository facilities for the Slovenian nuclear power plant Krsko are presented. The MCNP4C code was deployed to model and assess the neutron multiplication parameters of pool-based stor

  2. Criticality safety and sensitivity analyses of PWR spent nuclear fuel repository facilities

    NARCIS (Netherlands)

    Maucec, M; Glumac, B

    2005-01-01

    Monte Carlo criticality safety and sensitivity calculations of pressurized water reactor (PWR) spent nuclear fuel repository facilities for the Slovenian nuclear power plant Krsko are presented. The MCNP4C code was deployed to model and assess the neutron multiplication parameters of pool-based stor

  3. Identification of dose-reduction techniques for BWR and PWR repetitive high-dose jobs

    Energy Technology Data Exchange (ETDEWEB)

    Dionne, B.J.; Baum, J.W.

    1984-01-01

    As a result of concern about the apparent increase in collective radiation dose to workers at nuclear power plants, this project will provide information to industry in preplanning for radiation protection during maintenance operations. This study identifies Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) repetitive jobs, and respective collective dose trends and dose reduction techniques. 3 references, 2 tables. (ACR)

  4. Cool visitors

    CERN Multimedia

    2006-01-01

    Pictured, from left to right: Tim Izo (saxophone, flute, guitar), Bobby Grant (tour manager), George Pajon (guitar). What do the LHC and a world-famous hip-hop group have in common? They are cool! On Saturday, 1st July, before their appearance at the Montreux Jazz Festival, three members of the 'Black Eyed Peas' came on a surprise visit to CERN, inspired by Dan Brown's Angels and Demons. At short notice, Connie Potter (Head of the ATLAS secretariat) organized a guided tour of ATLAS and the AD 'antimatter factory'. Still curious, lead vocalist Will.I.Am met CERN physicist Rolf Landua after the concert to ask many more questions on particles, CERN, and the origin of the Universe.

  5. Assessment of void swelling in austenitic stainless steel PWR core internals.

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H. M.; Energy Technology

    2006-01-31

    As many pressurized water reactors (PWRs) age and life extension of the aged plants is considered, void swelling behavior of austenitic stainless steel (SS) core internals has become the subject of increasing attention. In this report, the available database on void swelling and density change of austenitic SSs was critically reviewed. Irradiation conditions, test procedures, and microstructural characteristics were carefully examined, and key factors that are important to determine the relevance of the database to PWR conditions were evaluated. Most swelling data were obtained from steels irradiated in fast breeder reactors at temperatures >385 C and at dose rates that are orders of magnitude higher than PWR dose rates. Even for a given irradiation temperature and given steel, the integral effects of dose and dose rate on void swelling should not be separated. It is incorrect to extrapolate swelling data on the basis of 'progressive compounded multiplication' of separate effects of factors such as dose, dose rate, temperature, steel composition, and fabrication procedure. Therefore, the fast reactor data should not be extrapolated to determine credible void swelling behavior for PWR end-of-life (EOL) or life-extension conditions. Although the void swelling data extracted from fast reactor studies is extensive and conclusive, only limited amounts of swelling data and information have been obtained on microstructural characteristics from discharged PWR internals or steels irradiated at temperatures and at dose rates comparable to those of a PWR. Based on this relatively small amount of information, swelling in thin-walled tubes and baffle bolts in a PWR is not considered a concern. As additional data and relevant research becomes available, the newer results should be integrated with existing data, and the worthiness of this conclusion should continue to be scrutinized. PWR baffle reentrant corners are the most likely location to experience high swelling

  6. Studies on advanced water-cooled reactors beyond generation Ⅲ for power generation

    Institute of Scientific and Technical Information of China (English)

    CHENG Xu

    2007-01-01

    China's ambitious nuclear power program motivates the country's nuclear community to develop advanced reactor concepts beyond generation Ⅲ to ensure a long-term, stable, and sustainable development of nuclear power. The paper discusses some main criteria for the selection of future water-cooled reactors by considering the specific Chinese situation. Based on the suggested selection criteria, two new types of water-cooled reactors are recommended for future Chinese nuclear power generation. The high conversion pressurized water reactor utilizes the present PWR technology to a large extent. With a conversion ratio of about 0.95, the fuel utilization is increased about 5 times. This significantly improves the sustainability of fuel resources. The supercritical water-cooled reactor has favorable features in economics,sustainability and technology availability. It is a logical extension of the generation Ⅲ PWR technology in China.The status of international R&D work is reviewed. A new supercritieal water-cooled reactor (SCWR) core structure (the mixed reactor core) and a new fuel assembly design (two-rows FA) are proposed. The preliminary analysis using a coupled neutron-physics/thermal-hydranlics method is carded out. It shows good feasibility for the new design proposal.

  7. Simulation of Desiccant Cooling

    Directory of Open Access Journals (Sweden)

    Kamaruddin A.

    2017-06-01

    Full Text Available Desiccant cooling system has been an attractive topic for study lately, due to its environmentally friendly nature. It also consume less electricity and capable to be operated without refrigerant. A simulation study was conducted using 1.5 m long ducting equipped with one desiccant wheel, one sensible heat exchanger wheel, one evaporative cooling chamber and two blowers and one electric heater. The simulation study used 8.16 m/s primary air, the drying coefficient from desiccant wheel, k1=2.1 (1/s, mass transfer coefficient in evaporative cooling, k2=1.2 kg vapor/s, heat transfer coefficient in desiccant wheel, h1=4.5 W/m2 oC, and heat transfer coefficient in sensible heat exchanger wheel h2= 4.5 W/m2 oC. The simulation results show that the final temperature before entering into the air conditioning room was 25 oC and RH of 65 %, were in accordance with the Indonesian comfort index.

  8. Development of a Safeguards Verification Method and Instrument to Detect Pin Diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies Phase I Study

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Y S; Sitaraman, S

    2008-12-24

    A novel methodology to detect diversion of spent fuel from Pressurized Water Reactors (PWR) has been developed in order to address a long unsolved safeguards verification problem for international safeguards community such as International Atomic Energy Agency (IAEA) or European Atomic Energy Community (EURATOM). The concept involves inserting tiny neutron and gamma detectors into the guide tubes of a spent fuel assembly and measuring the signals. The guide tubes form a quadrant symmetric pattern in the various PWR fuel product lines and the neutron and gamma signals from these various locations are processed to obtain a unique signature for an undisturbed fuel assembly. Signatures based on the neutron and gamma signals individually or in a combination can be developed. Removal of fuel pins from the assembly will cause the signatures to be visibly perturbed thus enabling the detection of diversion. All of the required signal processing to obtain signatures can be performed on standard laptop computers. Monte Carlo simulation studies and a set of controlled experiments with actual commercial PWR spent fuel assemblies were performed and validated this novel methodology. Based on the simulation studies and benchmarking measurements, the methodology developed promises to be a powerful and practical way to detect partial defects that constitute 10% or more of the total active fuel pins. This far exceeds the detection threshold of 50% missing pins from a spent fuel assembly, a threshold defined by the IAEA Safeguards Criteria. The methodology does not rely on any operator provided data like burnup or cooling time and does not require movement of the fuel assembly from the storage rack in the spent fuel pool. A concept was developed to build a practical field device, Partial Defect Detector (PDET), which will be completely portable and will use standard radiation measuring devices already in use at the IAEA. The use of the device will not require any information provided

  9. Simulation model and methodology for calculating the damage by internal radiation in a PWR reactor; Modelo de simulacion y metodologia para el calculo del dano por irradiacion en los internos de un reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.; Benito Hernandez, M.; Barreira Pereira, P.

    2012-07-01

    This study involves the development of the methodology and three-dimensional models to estimate the damage to the vessel internals of a commercial PWR reactor from irradiation history of operating cycles.

  10. The Power-weakness Ratios (PWR as a Journal Indicator: Testing the “Tournaments” Metaphor in Citation Impact Studies

    Directory of Open Access Journals (Sweden)

    Loet Leydesdorff

    2016-09-01

    Full Text Available Purpose: Ramanujacharyulu developed the Power-weakness Ratio (PWR for scoring tournaments. The PWR algorithm has been advocated (and used for measuring the impact of journals. We show how such a newly proposed indicator can empirically be tested. Design/methodology/approach: PWR values can be found by recursively multiplying the citation matrix by itself until convergence is reached in both the cited and citing dimensions; the quotient of these two values is defined as PWR. We study the effectiveness of PWR using journal ecosystems drawn from the Library and Information Science (LIS set of the Web of Science (83 journals as an example. Pajek is used to compute PWRs for the full set, and Excel for the computation in the case of the two smaller sub-graphs: (1 JASIST+ the seven journals that cite JASIST more than 100 times in 2012; and (2 MIS Quart+ the nine journals citing this journal to the same extent. Findings: A test using the set of 83 journals converged, but did not provide interpretable results. Further decomposition of this set into homogeneous sub-graphs shows that—like most other journal indicators—PWR can perhaps be used within homogeneous sets, but not across citation communities. We conclude that PWR does not work as a journal impact indicator; journal impact, for example, is not a tournament. Research limitations: Journals that are not represented on the “citing” dimension of the matrix—for example, because they no longer appear, but are still registered as “cited” (e.g. ARIST—distort the PWR ranking because of zeros or very low values in the denominator. Practical implications: The association of “cited” with “power” and “citing” with “weakness” can be considered as a metaphor. In our opinion, referencing is an actor category and can be Metaphor in Citation Impact Studies in terms of behavior, whereas “citedness” is a property of a document with an expected dynamics very different from that of

  11. Radiative heat transfer modelling in a PWR severe accident sequence

    Energy Technology Data Exchange (ETDEWEB)

    Magali Zabiego; Florian Fichot [Institut de Radioprotection et de Surete Nucleaire - BP 3 - 13115 Saint-paul-Lez-Durance (France); Pablo Rubiolo [Westinghouse Science and Technology - 1344 Beulah Road - Pittsburgh - PA 15235 (United States)

    2005-07-01

    a debris bed. In particular, an expression of the conductivity was established in cells in which remaining cylinders and debris particles coexist. In the present document, after a recall of the main lines of the modelling, an application to a reactor sequence is proposed. A severe accident transient with core degradation is simulated. The radiative transfer model is shown to behave properly and to smoothly calculate the transitions between the successive core configurations. A comparison with the more classical Hottel method shows that the present model gives a better prediction of the degradation progression owing to a more accurate estimate of the radial heat transfers. References: [1] M. Zabiego et al., ICARE/CATHARE V1: application to a PWR 900 MWe severe accident sequence, SARJ, Tokyo, 1999; [2] M. Zabiego, F. Fichot, P. Rubiolo Transfert radiatif lors d'une sequence accidentelle dans un coeur de Reacteur a Eau sous Pression, Congres Francais de Thermique, SFT 2004, Presqu'ile de Giens, 25-28 mai 2004. (authors)

  12. New cooling regulation technology of secondary cooling station in DCS

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Xuan; Yan, Jun-wei; Zhu, Dong-sheng; Liu, Fei-long; Lei, Jun-xi [The Key Lab of Enhanced Heat Transfer and Energy Conservation of Ministry of Education, School of Chemical and Energy Engineering, South China University of Technology, Guangzhou 510641 (China); Liang, Lie-quan [The Key Lab of E-Commerce Market Application Technology of Guangdong Province, Guangdong University of Business Studies, Guangzhou 510320 (China)

    2008-07-01

    In this paper, a kind of new control technology of secondary cooling station (constant flow rate/variable temperature difference) in district cooling system (DCS) is proposed in view of serial consequences including low efficiency and high operating cost caused by low temperature of supply water in DCS. This technology has been applied in DCS of Guangzhou University City. The result has already indicated that such technology can increase the supply and return temperatures of buildings, return water temperature of primary side in the plate heat exchanger unit, moreover, the efficiency of both the chiller and the whole system are improved significantly. (author)

  13. Accident analysis of heavy water cooled thorium breeder reactor

    Science.gov (United States)

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

    2015-04-01

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

  14. Propagation Limits of High Pressure Cool Flames

    Science.gov (United States)

    Ju, Yiguang

    2016-11-01

    The flame speeds and propagation limits of premixed cool flames at elevated pressures with radiative heat loss are numerically modelled using dimethyl ether mixtures. The primary focus is paid on the effects of pressure, mixture dilution, flame size, and heat loss on cool flame propagation. The results showed that cool flames exist on both fuel lean and fuel rich sides and thus dramatically extend the lean and rich flammability limits. There exist three different flame regimes, hot flame, cool flame, and double flame. A new flame flammability diagram including both cool flames and hot flames is obtained at elevated pressure. The results show that pressure significantly changes cool flame propagation. It is found that the increases of pressure affects the propagation speeds of lean and rich cool flames differently due to the negative temperature coefficient effect. On the lean side, the increase of pressure accelerates the cool flame chemistry and shifts the transition limit of cool flame to hot flame to lower equivalence ratio. At lower pressure, there is an extinction transition from hot flame to cool flame. However, there exists a critical pressure above which the cool flame to hot flame transition limit merges with the lean flammability limit of the hot flame, resulting in a direct transition from hot flame to cool flame. On the other hand, the increase of dilution reduces the heat release of hot flame and promotes cool flame formation. Moreover, it is shown that a smaller flame size and a higher heat loss also extend the cool flame transition limit and promote cool flame formation.

  15. Water Reflooding Effectiveness Assessment for 1 000 MWe PWR under Severe Accident Condition%百万千瓦级压水堆严重事故后再注水的有效性评价

    Institute of Scientific and Technical Information of China (English)

    胡啸; 黄挺; 裴杰; 陈炼

    2015-01-01

    根据现有的设计资料,使用一体化严重事故分析程序 MELCOR1.8.6建立了核电厂一、二回路系统,非能动堆芯冷却系统和安全壳系统的模型,并模拟冷段2英寸(5.08 cm)小破口叠加重力注入失效的严重事故发生后,将冷却剂注入堆芯的情形,分析其对严重事故进程的缓解能力。本文选取3个严重事故的不同阶段,将冷却剂分别以小流量(10 kg/s)、中流量(50 kg/s)和大流量(200 kg/s)的速率注入堆芯,通过比较氢气产生量、堆芯放射性产生量及堆芯温度等数据来评估在严重事故不同阶段再注水的可行性。结果表明:在堆芯损伤初期,可认为10 kg/s以上的流量足以冷却百万千瓦级事故安全。而当严重事故发展到堆芯开始坍塌阶段,200 kg/s的注水流量可认为是基本可行的,而小于此流量的注水应慎重考虑。%The MELCOR1.8.6 code was applied to a severe accident model of a 1 000 MWe PWR which includes primary system,secondary system,passive core cool-ing system and containment system.For the transient case,a small break LOCA with 2 inch (5.08 cm)break at the cold leg concurrent with failure of gravity injection was selected.After the core was damaged due to the failure of gravity inj ection,it was assumed that the coolant was inj ected into the pressure vessel,and then the water reflooding effectiveness was evaluated and analyzed.In this calculation,the coolant injection into reactor core with the small (10 kg/s),medium (50 kg/s)and large (200 kg/s)mass flow rates respectively at 3 different time stages of the severe accident was simulated.The effectiveness of water reflooding was assessed through hydrogen production,radioactive materials released from core,and core temperature.The results show that the mass flow rate above 10 kg/s is believed to be efficient for cooling a 1 000 MWe reactor at the beginning of core damage.However,with the accident devel-oping to core relocation,a large mass flow

  16. Study on stress corrosion of the zone affected by the AISI 316L steel heat under PWR reactor environment at 325 deg Celsius; Estudo da corrosao sob tensao da zona afetada pelo calor do aco AISI 316L em ambiente de reator PWR a 325 deg C

    Energy Technology Data Exchange (ETDEWEB)

    Satler Filho, Luiz F.; Schvartzman, Monica M.A.M.; Quinan, Marco A.D.; Soares, Antonio E.G., E-mail: aegs@cdtn.b, E-mail: fernandosatler@yahoo.com.b, E-mail: quinanm@cdtn.b, E-mail: monicas@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Lima, Luciana I.L., E-mail: lill@cdtn.b [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil)

    2009-07-01

    This paper evaluates the stress corrosion susceptibility of the HAZ (heat affected zone) of the AISI 316L stainless steel of a dissimilar welding done between the ASTM A-508 steel and the AISI 316L steel, using a nickel alloy, under a chemical environment similar to the PWR (Pressurized Water Reactor) nuclear reactor primary circuit. The nickel 82 and 182 alloys were used in the GTAW (Gas Tungsten Arc Welding) and SMAW (Shielded Metal Arc Welding) processes respectively. The test at slow deformation - SSRT (Slow Strain Rate Test) was applied, using a deformation rate of 3x10{sup -7} s{sup -1}, at a temperature of 325 degree Celsius and pressure of 12.5 MPa. The susceptibility under tress corrosion evaluation was performed comparing the resistance limit, the total deformation and the fracture time obtained at the inert medium (nitrogen) and at the PWR medium. Also, the fracture surfaces were observed under a scanning electron microscope, verifying the fragile fracture regions

  17. Experiment data report for LOFT nonnuclear Test L1-4. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Batt, D. L.

    1977-07-01

    Test L1-4 was the fourth in a series of five nonnuclear isothermal blowdown tests conducted by the Loss of Fluid Test (LOFT) Program. Test L1-4 was the first Nuclear Regulatory Commission standard problem (International Problem No. 5 and U.S. Problem No. 7) experiment conducted at LOFT. Data from this test will be compared with predictions generated by the standard problem participants. For this test the LOFT Facility was configured to simulate a loss-of-coolant accident in a large pressurized water reactor resulting from a 200% double-ended offset shear break in a cold leg of the primary coolant system. A hydraulic core simulator assembly was installed in place of the nuclear core. The initial conditions in the primary coolant system intact loop were temperature at 279/sup 0/C, gauge pressure at 15.65 MPa, and intact loop flow at 268.4 kg/s. During system depressurization into a simulated containment, emergency core cooling water was injected into the primary coolant system cold leg to provide data on the effects of emergency core cooling on system thermalhydraulic response.

  18. Analysis of WWER-440 and PWR RPV welds surveillance data to compare irradiation damage evolution

    Energy Technology Data Exchange (ETDEWEB)

    Debarberis, L. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands)]. E-mail: luigi.debarberis@cec.eu.int; Acosta, B. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands)]. E-mail: beatriz.acosta-iborra@jrc.nl; Zeman, A. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Sevini, F. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Ballesteros, A. [Tecnatom, Avd. Montes de Oca 1, San Sebasitan de los Reyes, E-28709 Madrid (Spain); Kryukov, A. [Russian Research Centre Kurchatov Institute, Kurchatov Square 1, 123182 Moscow (Russian Federation); Gillemot, F. [AEKI Atomic Research Institute, Konkoly Thege M. ut 29-33, 1121 Budapest (Hungary); Brumovsky, M. [NRI, Nuclear Research Institute, Husinec-Rez 130, 25068 Rez (Czech Republic)

    2006-04-15

    It is known that for Russian-type and Western water reactor pressure vessel steels there is a similar degradation in mechanical properties during equivalent neutron irradiation. Available surveillance results from WWER and PWR vessels are used in this article to compare irradiation damage evolution for the different reactor pressure vessel welds. The analysis is done through the semi-mechanistic model for radiation embrittlement developed by JRC-IE. Consistency analysis with BWR vessel materials and model alloys has also been performed within this study. Globally the two families of studied materials follow similar trends regarding the evolution of irradiation damage. Moreover in the high fluence range typical of operation of WWER the radiation stability of these vessels is greater than the foreseen one for PWR.

  19. Eddy current NDT: a suitable tool to measure oxide layer thickness in PWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Alencar, Donizete A.; Silva Junior, Silverio F. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil)], e-mail: daa@cdtn.br, e-mail: silvasf@cdtn.br; Vieira, Andre L.P.S. [Industrias Nucleares do Brasil (INB S.A.), Resende, RJ (Brazil). Fabrica de Combustivel Nuclear], e-mail: andre@inb.gov.br; Soares, Adolpho [Technotest Consultoria e Acessoria Ltda., Belo Horizonte, MG (Brazil)], e-mail: adolpho@technotest.com.br

    2009-07-01

    Eddy current is a nondestructive test (NDT) widely used in industry to support integrity analysis of components and equipment. In the nuclear area it is frequently applied to inspect tubes installed in tube exchangers, such as steam generators and condensers in PWR plants, as well as turbine blades. Adequately assisted by means of robotic devices, that inspection method has been pointed as a suitable tool to perform accurate oxide layer thickness measurements in PWR fuel rods. This paper shows some theoretical aspects and physical operating principles of the inspection method, as well as test probes construction details, and the calibration reference standards fabrication processes. Furthermore, some data, experimentally obtained at INB laboratories and other technical information obtained from TECNATOM S.A. are presented, showing the accuracy and efficacy of such NDT method. (author)

  20. PENGARUH KONDISI ATMOSFERIK TERHADAP PERHITUNGAN PROBABILISTIK DAMPAK RADIOLOGI KECELAKAAN PWR 1000-MWe

    Directory of Open Access Journals (Sweden)

    Pande Made Udiyani

    2015-10-01

    Full Text Available ABSTRAK PENGARUH KONDISI ATMOSFERIK TERHADAP PERHITUNGAN PROBABILISTIK DAMPAK RADIOLOGI KECELAKAAN PWR 1000-MWe.  Perhitungan dampak kecelakaan radiologi terhadap lepasan produk fisi akibat kecelakaan potensial yang mungkin terjadi di Pressurized Water Reactor (PWR diperlukan secara probabilistik. Mengingat kondisi atmosfer sangat berperan terhadap dispersi radionuklida di lingkungan, dalam penelitian ini akan dianalisis pengaruh kondisi atmosferik terhadap perhitungan probabilistik dari konsekuensi kecelakaan reaktor.  Tujuan penelitian adalah melakukan analisis terhadap pengaruh kondisi atmosfer berdasarkan model data input meteorologi terhadap dampak radiologi kecelakaan PWR 1000-MWe yang disimulasikan pada tapak yang mempunyai kondisi meteorologi yang berbeda. Simulasi menggunakan program PC-Cosyma dengan moda perhitungan probabilistik, dengan data input meteorologi yang dieksekusi secara cyclic dan stratified, dan disimulasikan di Tapak Semenanjung Muria dan Pesisir Serang. Data meteorologi diambil setiap jam untuk jangka waktu satu tahun. Hasil perhitungan menunjukkan bahwa frekuensi kumulatif  untuk model input yang sama untuk Tapak pesisir Serang lebih tinggi dibandingkan dengan Semenanjung Muria. Untuk tapak yang sama, frekuensi kumulatif model input cyclic lebih tinggi dibandingkan model stratified. Model cyclic memberikan keleluasan dalam menentukan tingkat ketelitian perhitungan dan tidak membutuhkan data acuan dibandingkan dengan model stratified. Penggunaan model cyclic dan stratified melibatkan jumlah data yang besar dan pengulangan perhitungan  akan meningkatkan  ketelitian nilai-nilai statistika perhitungan. Kata kunci: dampak kecelakaan, PWR 1000-MWe,  probabilistik,  atmosferik, PC-Cosyma   ABSTRACT THE INFLUENCE OF ATMOSPHERIC CONDITIONS TO PROBABILISTIC CALCULATION OF IMPACT OF RADIOLOGY ACCIDENT ON PWR-1000MWe. The calculation of the radiological impact of the fission products releases due to potential accidents

  1. DOMINO: A fast 3D cartesian discrete ordinates solver for reference PWR simulations and SPN validation

    Energy Technology Data Exchange (ETDEWEB)

    Courau, T.; Moustafa, S.; Plagne, L.; Poncot, A. [EDF R and D, 1, Av du General de Gaulle, F92141 Clamart cedex (France)

    2013-07-01

    As part of its activity, EDF R and D is developing a new nuclear core simulation code named COCAGNE. This code relies on DIABOLO, a Simplified PN (SPN) method to compute the neutron flux inside the core for eigenvalue calculations. In order to assess the accuracy of SPN calculations, we have developed DOMINO, a new 3D Cartesian SN solver. The parallel implementation of DOMINO is very efficient and allows to complete an eigenvalue calculation involving around 300 x 10{sup 9} degrees of freedom within a few hours on a single shared-memory supercomputing node. This computation corresponds to a 26-group S{sub 8} 3D PWR core model used to assess the SPN accuracy. At the pin level, the maximal error for the SP{sub 5} DIABOLO fission production rate is lower than 0.2% compared to the S{sub 8} DOMINO reference for this 3D PWR core model. (authors)

  2. EPRI PWR Safety and Relief Value Test Program: safety and relief valve test report

    Energy Technology Data Exchange (ETDEWEB)

    1982-12-01

    A safety and relief valve test program was conducted by EPRI for a group of participating PWR utilities to respond to the USNRC recommendations documented in NUREG 0578 Section 2.1.2, and as clarified in NUREG 0737 Item II.D.1.A. Seventeen safety and relief valves representative of those utilized in or planned for use in participating domestic PWR's were tested under the full range of selected test conditions. This report contains a listing of the selected test valves and the corresponding as tested test matrices, valve performance data and principal observations for the tested safety and relief valves. The information contained in this report may be used by the participating utilities in developing their response to the above mentioned USNRC recommendations.

  3. Study for identification of control rod drops in PWR reactors at any burnup step

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Thiago J.; Martinez, Aquilino S.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail: tsouza@nuclear.ufrj.br, E-mail: aquilino@lmp.ufrj.br, E-mail: canedo@lmp.ufrj.br, E-mail: alessandro@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    The control rod drop event in PWR reactors induces an unsafe operating condition. Therefore, in a scenario of a control rod drop is important to quickly identify the rod to minimize undesirable effects. The objective of this work is to develop an on-line method for identification of control rod drop in PWR reactors. The method consists on the construction of a tool that is based on the ex-core detector responses. Therefore, it is proposed to recognize patterns in the neutron ex-core detectors responses and thus to identify on-line a control rod drop in the core during the reactor operation. The results of the study, as well as the behavior of the detector responses, demonstrated the feasibility of this method. (author)

  4. Proving test on the seismic reliability of nuclear power plant: PWR reactor containment vessel

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Hiroshi; Yoshikawa, Teiichi; Ohno, Tokue; Yoshikawa, Eiji.

    1989-01-01

    Seismic reliability proving tests of nuclear power plant facilities are carried out by the Nuclear Power Engineering Test Center, using the large-scale, high-performance vibration table of Tadotsu Engineering Laboratory, and sponsored by the Ministry of International Trade and Industry. In 1982, the seismic reliability proving test of a PWR containment vessel was conducted using a test component of reduced scale 1/3.7. As a result of this test, the test component proved to have structural soundness against earthquakes, and at the same time its stable function was proved by leak tests which were carried out before and after the vibration test. In 1983, the detailed analysis and evaluation of these test results were carried out, and the analysis methods for evaluating strength against earthquakes were established. The seismic analysis and evaluation on the actual containment vessel were then performed using these analysis methods, and the safety and reliability of the PWR reactor containment vessel were confirmed.

  5. Colloids: a review of current knowledge with a view to application to phenomena of transportation within PWR; Colloides: point de vue sur les connaissances actuelles en vue d`une application aux phenomenes de transport dans les REP

    Energy Technology Data Exchange (ETDEWEB)

    Guinard, L.

    1996-12-31

    In an attempt to minimise dosimetry within the primary circuit of PWR units, research is being carried out into understanding the phenomena of transportation and deposition of corrosion products. It is therefore desirable to known the form of these corrosion products and the laws governing this form. It is generally considered that they are in soluble or particulate form. A third starts with a general presentation of colloids and goes on to define points which are useful, both on a theoretical and experimental level, in terms of application to phenomena of transportation within PWRs. (author). 69 refs., 30 figs., 6 tabs., 3 appends.

  6. MELCOR model for an experimental 17x17 spent fuel PWR assembly.

    Energy Technology Data Exchange (ETDEWEB)

    Cardoni, Jeffrey

    2010-11-01

    A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

  7. Effect of proton irradiation on irradiation assisted stress corrosion cracking in PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Han Ok; Hwang, Mi Jin; Kim, Sung Woo; Hwang, Seong Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Irradiation assisted stress corrosion cracking (IASCC) involves the cracking and failure of materials under irradiation environment in nuclear power plant water environment. The major factors and processes governing an IASCC are suggested by others. The IASCC of the reactor core internals due to the material degradation and the water chemistry change has been reported in high stress stainless steel components, such as fuel elements (Boiling Water Reactors) in the 1960s, a control rod in the 1970s, and a baffle former bolt in recent years of light water reactors (Pressurized Water Reactors). Many irradiated stainless steels that are resistant to inergranular cracking in 288 .deg. C argon are susceptible to IG cracking in the simulated BWR environment at the same temperature. Under the circumstances, a lot works have been performed on IASCC in BWR. Recent efforts have been devoted to investigate an IASCC in a PWR, but the mechanism in a PWR is not fully understood yet as compared with that in a BWR owing to a lack of data from laboratories and fields. Therefore, it is strongly necessary to review and analyze recent researches of an IASCC in both BWR and PWR for establishing a proactive management technology for the IASCC of core internals in Korean PWRs. The objective of this research to find IASCC behavior of proton irradiated 316 stainless steels in a high-temperature water chemistry environment. The IASCC initiation susceptibility on 1, 3, 5 DPA proton irradiated 316 austenite stainless steel was evaluated in PWR environment. SCC area ratio on the fracture surface was similar regardless of irradiation level. Total crack length on the irradiated surface increases in order of specimen 1, 3, 5 DPA. The total crack length at the side surface is a better measure in evaluating IASCC initiation susceptibility for proton-irradiated samples.

  8. Anti -corrosion Effect of ETA on Materials in Secondary Loop of PWR

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    In the world, over sixty percent of nuclear power plant have used advanced amunes ETA(Ethanolamine) as pH control agent in secondary loop of PWR. There are eighty percent of nuclear powerplants using ETA in USA. The corrosion of materials in steam generator (SG) tube and secondary looppower water reactor have been inhibited, the life of SG and the economics of the plant are increasedbecause of using ETA.

  9. Chemical and radiochemical specifications - PWR power plants; Specifications chimiques et radiochimiques - Centrales REP

    Energy Technology Data Exchange (ETDEWEB)

    Stutzmann, A. [Electricite de France (EDF), 93 - Saint-Denis (France)

    1997-07-01

    Published by EDF this document gives the chemical specifications of the PWR (Pressurized Water Reactor) nuclear power plants. Among the chemical parameters, some have to be respected for the safety. These parameters are listed in the STE (Technical Specifications of Exploitation). The values to respect, the analysis frequencies and the time states of possible drops are noticed in this document with the motion STE under the concerned parameter. (A.L.B.)

  10. Proof test on thermal and hydraulic design reliability of Japanese PWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Mamoru (Univ. of Tokyo (Japan)); Inoue, Akira (Tokyo Institute of Technology (Japan)); Miyazaki, Keiji (Osaka Univ. (Japan)); Abeta, Sadaaki (Mitsubishi, Tokyo (Japan)); Hori, Keiichi (Mitsubishi, Hyogo (Japan)); Mukasa, Tomio; Oishi, Masao; Aoki, Toshimasa; Makihara, Yoshiaki

    1990-01-01

    A series of departure from nucleate boiling (DNB) tests for pressurized water reactors (PWRs) was performed at the Nuclear Power Engineering Test Center. The objective was to prove the reliability of fuel assembly design by confirming the thermal margin of heat transfer. The present method for evaluating the DNB ratio in a Japanese 17 x 17 PWR core is adequate according to the newly obtained DNB test data.

  11. EDF/CIDEN - ONECTRA: PWR decontamination; EDF/CIDEN - ONECTRA: assainissement REP

    Energy Technology Data Exchange (ETDEWEB)

    Fayolle, P. [EDFICIDEN, 35-37, rue Louis Guerin - B.P. 21212, 69611 Villeurbanne Cedex (France); Orcel, H. [ONECTRA, ZA les Tomples BP45, 26701 Pierrelatte Cedex (France); Wertz, L. [ONECTRA, Le Britannia, Allee C, 20 Bd Eugene Deruelle, 69432 Lyon Cedex 03 (France)

    2010-07-01

    In the context of PWR circuit renewal (expected in 2011) and their decontamination, an analysis of data coming from cartography and on site decontamination measurements as well as from premise modelling by means of the PANTHERE radioprotection code, is presented. Several French PWRs have been studied. After a presentation of code principles and operation, the authors discuss the radiological context of a workstation, and give an assessment of the annual dose associated with maintenance operations with or without decontamination

  12. PWR ENDF/B-VII cross-section libraries for ORIGEN-ARP

    Energy Technology Data Exchange (ETDEWEB)

    McGraw, C. [Dept. of Nuclear Engineering, Texas A and M Univ., 3133 TAMU, College Station, TX 77843-3133 (United States); Ilas, G. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6172 (United States)

    2012-07-01

    New pressurized water reactor (PWR) cross-section libraries were generated for use with the ORIGEN-ARP depletion sequence in the SCALE nuclear analysis code system. These libraries are based on ENDF/B-VII nuclear data and were generated using the two-dimensional depletion sequence, TRITON/NEWT, in SCALE 6.1. The libraries contain multiple burnup-dependent cross-sections for seven PWR fuel designs, with enrichments ranging from 1.5 to 6 wt% {sup 235}U. The burnup range has been extended from the 72 GWd/MTU used in previous versions of the libraries to 90 GWd/MTU. Validation of the libraries using radiochemical assay measurements and decay heat measurements for PWR spent fuel showed good agreement between calculated and experimental data. Verification against detailed TRITON simulations for the considered assembly designs showed that depletion calculations performed in ORIGEN-ARP with the pre-generated libraries provide similar results as obtained with direct TRITON depletion, while greatly reducing the computation time. (authors)

  13. Calculation of the radionuclides in PWR spent fuel samples for SFR experiment planning.

    Energy Technology Data Exchange (ETDEWEB)

    Naegeli, Robert Earl

    2004-06-01

    This report documents the calculation of radionuclide content in the pressurized water reactor (PWR) spent fuel samples planned for use in the Spent Fuel Ratio (SPR) Experiments at Sandia National Laboratories, Albuquerque, New Mexico (SNL) to aid in experiment planning. The calculation methods using the ORIGEN2 and ORIGEN-ARP computer codes and the input modeling of the planned PWR spent fuel from the H. B. Robinson and the Surry nuclear power plants are discussed. The safety hazards for the calculated nuclide inventories in the spent fuel samples are characterized by the potential airborne dose and by the portion of the nuclear facility hazard category 2 and 3 thresholds that the experiment samples would present. In addition, the gamma ray photon energy source for the nuclide inventories is tabulated to facilitate subsequent calculation of the direct and shielded dose rates expected from the samples. The relative hazards of the high burnup 72 gigawatt-day per metric ton of uranium (GWd/MTU) spent fuel from H. B. Robinson and the medium burnup 36 GWd/MTU spent fuel from Surry are compared against a parametric calculation of various fuel burnups to assess the potential for higher hazard PWR fuel samples.

  14. A highly heterogeneous 3D PWR core benchmark: deterministic and Monte Carlo method comparison

    Science.gov (United States)

    Jaboulay, J.-C.; Damian, F.; Douce, S.; Lopez, F.; Guenaut, C.; Aggery, A.; Poinot-Salanon, C.

    2014-06-01

    Physical analyses of the LWR potential performances with regards to the fuel utilization require an important part of the work dedicated to the validation of the deterministic models used for theses analyses. Advances in both codes and computer technology give the opportunity to perform the validation of these models on complex 3D core configurations closed to the physical situations encountered (both steady-state and transient configurations). In this paper, we used the Monte Carlo Transport code TRIPOLI-4®; to describe a whole 3D large-scale and highly-heterogeneous LWR core. The aim of this study is to validate the deterministic CRONOS2 code to Monte Carlo code TRIPOLI-4®; in a relevant PWR core configuration. As a consequence, a 3D pin by pin model with a consistent number of volumes (4.3 millions) and media (around 23,000) is established to precisely characterize the core at equilibrium cycle, namely using a refined burn-up and moderator density maps. The configuration selected for this analysis is a very heterogeneous PWR high conversion core with fissile (MOX fuel) and fertile zones (depleted uranium). Furthermore, a tight pitch lattice is selcted (to increase conversion of 238U in 239Pu) that leads to harder neutron spectrum compared to standard PWR assembly. In these conditions two main subjects will be discussed: the Monte Carlo variance calculation and the assessment of the diffusion operator with two energy groups for the core calculation.

  15. Evaluation of the RELAP4/MOD6 thermal-hydraulic code. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Haigh, W.S.; Margolis, S.G.; Rice, R.E.

    1978-01-01

    The NRC RELAP4/MOD6 computer code was recently released to the public for use in thermal-hydraulic analysis. This code has a unique new capability permitting analysis of both the blowdown and reflood portions of a postulated pressurized water reactor (PWR) loss-of-coolant accident (LOCA). A principal code evaluation objective is to assess the accuracy of the code for computing LOCA behavior over a wide range of system sizes and scaling concepts. The scales of interest include all LOCA experiments and will ultimately encompass full-sized PWR systems for which no experiments or data are available. Quantitative assessment of the accuracy of the code when it is applied to large PWR systems is still in the future. With RELAP4/MOD6, however, a technique has been demonstrated for using results derived from small-scale blowdown and reflood experiments to predict the accuracy of calculations for similar experiments of significantly different scale or component size. This demonstration is considered a first step in establishing confidence levels for the accuracy of calculations of a postulated LOCA.

  16. Correlation Equations of Heat Transfer in Nanofluid Al2O3-Water as Cooling Fluid in a Rectangular Sub Channel Based CFD Code

    Directory of Open Access Journals (Sweden)

    Anwar Ilmar Ramadhan

    2015-03-01

    Full Text Available Safety is a major concern in the design, operation and development of a nuclear reactor. One aspect of nuclear reactor safety factor is thermal-hydraulics aspect. In a PWR-type nuclear power plant has been used lighter fluid coolant is water or H2O. In this research, using nanofluid Al2O3-Water with volume fraction of (1%, (2% and also (3%, used as a cooling fluid in a nuclear reactor core with sub channel PWR fuel element rectangular arrangement. This research was carried out modeling of fuel elements are arranged rectangular, then performed numerical simulations using Computational Fluid Dynamics (CFD code. In order to obtain the characteristic pattern of flow velocity of each fluid, the fluid temperature distribution along the cylinder wall temperature distribution of the fuel element. Then analyzed the heat transfer in a nuclear reactor core with sub channel PWR fuel element rectangular arrangement, including heat transfer coefficient, Nusselt number (Nu, as well as heat transfer correlations. Heat transfer correlation for nanofluid Al2O3-Water (1%, (2% and also (3% proved to core of PWR nuclear reactor fuel element sub channel rectangular arrangement with the Reynolds number (Re is stretched, namely: 404 096

  17. Hybrid radiator cooling system

    Science.gov (United States)

    France, David M.; Smith, David S.; Yu, Wenhua; Routbort, Jules L.

    2016-03-15

    A method and hybrid radiator-cooling apparatus for implementing enhanced radiator-cooling are provided. The hybrid radiator-cooling apparatus includes an air-side finned surface for air cooling; an elongated vertically extending surface extending outwardly from the air-side finned surface on a downstream air-side of the hybrid radiator; and a water supply for selectively providing evaporative cooling with water flow by gravity on the elongated vertically extending surface.

  18. Development of Calculation Code for Fission Product and Corrosion Product in PWR’s Primary Loop

    Institute of Scientific and Technical Information of China (English)

    XU; Zhi-long; WAN; Hai-xia; SHAO; Jing; WU; Xiao-chun; LI; Long; LIU; Xing-min; KE; Guo-tu

    2015-01-01

    With the basis of study on generation,release and migration of fission product,calculation model for each of the above processes was developed,and calculation method for source term of PWR fission products was established.Study on source term of corrosion product in primary loop was been done.Based on the study of corrosion,

  19. Metallurgical and mechanical behaviours of PWR fuel cladding tube oxidised at high temperature; Comportements metallurqigue et mecanique des materiaux de gainage du combustible REP oxydes a haute temperature

    Energy Technology Data Exchange (ETDEWEB)

    Stern, A

    2007-12-15

    Zirconium alloys are used as cladding materials in Pressurized Water Reactors (PWR). As they are submitted to very extreme conditions, it is necessary to check their behaviour and especially to make sure they meet the safety criteria. They are therefore studied under typical in service-loadings but also under accidental loadings. In one of these accidental scenarios, called Loss of Coolant Accident (LOCA) the cladding temperature may increase above 800 C, in a steam environment, and decrease before a final quench of the cladding. During this temperature transient, the cladding is heavily oxidised, and the metallurgical changes lead to a decrease of the post quench mechanical properties. It is then necessary to correlate this drop in residual ductility to the metallurgical evolutions. This is the problem we want to address in this study: the oxidation of PWR cladding materials at high temperature in a steam environment and its consequences on post quench mechanical properties. As oxygen goes massively into the metallic part - a zirconia layer grows at the same time - during the high temperature oxidation, the claddings tubes microstructure shows three different phases that are the outer oxide layer (zirconia) and the inner metallic phases ({alpha}(O) and 'ex {beta}') - with various mechanical properties. In order to reproduce the behaviour of this multilayered material, the first part of this study consisted in creating samples with different - but homogeneous in thickness - oxygen contents, similar to those observed in the different phases of the real cladding. The study was especially focused on the {beta}-->{alpha} phase transformation upon cooling and on the resulting microstructures. A mechanism was proposed to describe this phase transformation. For instance, we conclude that for our oxygen enriched samples, the phase transformation kinetics upon cooling are ruled by the oxygen partitioning between the two allotropic phases. Then, these materials

  20. Study on water cooled high conversion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    As a part of study on advanced reactors for the future, conceptual design of high conversion water cooled reactors is being studied, aiming at the contribution to nuclear fuel cycle by the LWR technology, since the utilization of LWRs will extend over a long period of time . We are studying on the reactor core concepts for BWR and PWR reactor systems. As for BWR system, three types of reactor cores are investigating for three different design goals; long operation period, high conversion ratio and high applicability for the existing BWR system. In all the cases, we have obtained a fair prospect of a large core concept with a capacity of 1,000 MWe class having negative void reactivity coefficient. This study is a part of JAERI-JAPCO (Japan Atomic Power Company) cooperative studies. Various kinds of conceptual designs will be created until the end of FY 1999. The designs will be checked and reviewed at that time, then experimental studies on the realization of the concepts will start with further design works from FY 2000. (author)

  1. Dry cooling tower with water augmentation

    Energy Technology Data Exchange (ETDEWEB)

    Ireland, R.G.; Tramontini, V.N.

    1981-06-23

    An air cooling tower system is disclosed for condensing exhaust steam in power plants, that has water cooling augmentation to maintain the plant cooling capacity during high atmospheric temperature periods. The cooling tower includes a plurality of banks of brazed aluminum plate and fin type heat exchangers arranged in inverted ''v'' shaped sets. These heat exchangers cool ammonia used as the cooling fluid in the primary condenser for the power plant turbine exhaust steam. Each of these heat exchangers has a core consisting of a plurality of parallel aluminum plates spaced apart by fin assemblies that define a plurality of fluid passes. Approximately every other one of these passes has closed sides that open at the ends of the core to headers and define ammonia passes. The passes adjacent the ammonia passes are open at the sides and define air passes that permit the free flow of air transversely through the heat exchanger cores. An additional pass is provided adjacent every fourth one of the ammonia passes and these have closed sides and ends and define the passes for the cooling water. The water passes communicate at the bottom of the core with a water inlet manifold and at the top of the core with a water outlet manifold. The cooling tower system is designed so that at 55 degrees fahrenheit air temperatures or below, the cooling air alone will provide the necessary cooling for the ammonia to satisfy plant requirements. Above 55 degrees fahrenheit air temperature, cooling water from a separate water tank is pumped through the water passes to provide an additional cooling effect to maintain the design cooling capacity.

  2. Reactor Physics Assessment of Thick Silicon Carbide Clad PWR Fuels

    Science.gov (United States)

    2013-06-01

    Loss of Coolant Accident LWR Light Water Reactor MOX Mixed Oxide Fuel MTC Moderator Temperature Coefficient MWd/kgIHM Megawatt days per...working only with UO2 and UO2/PuO2 mixed oxide ( MOX ) fuels. 3.1 Studsvik Core Management Software CASMO-4E and SIMULATE-3 are the primary computational

  3. Estimation of droplets/wall heat transfer under LOCA conditions in a PWR; Estimation du transfert de chaleur gouttes/paroi en situation d'APRP pour un REP

    Energy Technology Data Exchange (ETDEWEB)

    GrAdeck, M.; Maillet, D. [CNRS UMR 7563 2, 54 - Vandoeuvre les Nancy (France); Lelong, F.; Seiler, N.; Repetto, G. [IRSN Cadarache, 13 - Saint Paul lez Durance (France)

    2009-07-01

    During a LOCA (Loss Of Coolant Accident) in a PWR, the fuel assemblies could be locally severely ballooned. The transient is ended by the injection of water initiated the safety system. The cooling of theses partially blocked fuel assemblies depends on the coolant flow characteristics in the blockage region. Most models for heat transfers concentrate on cooling of the ballooned walls by vapor convection. Since a two-phase mist flow occurs when reflooding, the possibility of additional cooling by direct liquid droplet impingement on the blockage surfaces must be investigated. As the temperature of the fuel assemblies is higher than the Leidenfrost temperature, the impact regime should be only the bouncing one. Up to now, no model of heat transfer of droplet impacts has been developed for that regime. As the coolability from droplet impacts must be modeled, an experimental program was proposed with droplets and wall characteristics (velocity, diameter, temperature) close to the LOCA ones. As the interaction between the droplet and the wall is very short (a few of ms), the estimation of the heat flux during the resident time of the droplet at the wall must be accurately designed. The purpose of this work is to show how such heat flux can be experimentally estimated used an adapted inverse heat conduction model. The final goal of the present collaboration between LEMTA and IRSN is to introduce the cooling model within NEPTUNE-CFD code, a joint project of CEA, EDF, AREVA and IRSN. (authors)

  4. Liquid-Cooled Garment

    Science.gov (United States)

    1977-01-01

    A liquid-cooled bra, offshoot of Apollo moon suit technology, aids the cancer-detection technique known as infrared thermography. Water flowing through tubes in the bra cools the skin surface to improve resolution of thermograph image.

  5. Cool Core Clusters from Cosmological Simulations

    CERN Document Server

    Rasia, E; Murante, G; Planelles, S; Beck, A M; Biffi, V; Ragone-Figueroa, C; Granato, G L; Steinborn, L K; Dolag, K

    2015-01-01

    We present results obtained from a set of cosmological hydrodynamic simulations of galaxy clusters, aimed at comparing predictions with observational data on the diversity between cool-core and non-cool-core clusters. Our simulations include the effects of stellar and AGN feedback and are based on an improved version of the Smoothed-Particle-Hydrodynamics code GADGET-3, which ameliorates gas mixing and better captures gas-dynamical instabilities by including a suitable artificial thermal diffusion. In this Letter, we focus our analysis on the entropy profiles, our primary diagnostic to classify the degree of cool-coreness of clusters, and on the iron profiles. In keeping with observations, our simulated clusters display a variety of behaviors in entropy profiles: they range from steadily decreasing profiles at small radii, characteristic of cool-core systems, to nearly flat core isentropic profiles, characteristic of non cool-core systems. Using observational criteria to distinguish between the two classes of...

  6. Design of a rapidly cooled cryogenic mirror

    Science.gov (United States)

    Plummer, Ron; Hsu, Ike

    1993-01-01

    The paper discusses the design, analysis, and testing of a rapidly cooled beryllium cryogenic mirror, which is the primary mirror in the four-element optical system for the Long Wavelength Infrared Advanced Technology Seeker. The mirror is shown to meet the requirement of five minutes for cooling to cryogenic operating temperature; it also maintains its optical figure and vacuum integrity and meets the nuclear specification. Results of a detailed thermal analysis on the mirror showed that, using nitrogen gas at 80 K as coolant, the front face of the mirror can be cooled from an initial temperature of 300 K to less than 90 K within five minutes. In a vacuum chamber, using liquid nitrogen as coolant, the mirror can be cooled to 80 K within 1.5 min. The mirror is well thermally insulated, so that it can be maintained at less than its operating temperature for a long time without active cooling.

  7. A neural networks based ``trip`` analysis system for PWR-type reactors; Um sistema de analise de ``trip`` em reatores PWR usando redes neuronais

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Antonio Carlos Pinto Dias

    1993-12-31

    The analysis short after automatic shutdown (trip) of a PWR-type nuclear reactor takes a considerable amount of time, not only because of the great number of variables involved in transients, but also the various equipment that compose a reactor of this kind. On the other hand, the transients`inter-relationship, intended to the detection of the type of the accident is an arduous task, since some of these accidents (like loss of FEEDWATER and station BLACKOUT, for example), generate transients similar in behavior (as cold leg temperature and steam generators mixture levels, for example). Also, the sequence-of-events analysis is not always sufficient for correctly pin point the causes of the trip. (author) 11 refs., 39 figs.

  8. Optimization of thermal efficiency of nuclear central power like as PWR; Otimizacao da eficiencia termica de uma usina nuclear do tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lapa, Nelbia da Silva

    2005-10-15

    The main purpose of this work is the definition of operational conditions for the steam and power conservation of Pressurized Water Reactor (PWR) plant in order to increase its system thermal efficiency without changing any component, based on the optimization of operational parameters of the plant. The thermal efficiency is calculated by a thermal balance program, based on conservation equations for homogeneous modeling. The circuit coefficients are estimated by an optimization tool, allowing a more realistic thermal balance for the plans under analysis, as well as others parameters necessary to some component models. With the operational parameter optimization, it is possible to get a level of thermal efficiency that increase capital gain, due to a better relationship between the electricity production and the amount of fuel used, without any need to change components plant. (author)

  9. Valve inlet fluid conditions for pressurizer safety and relief valves in Westinghouse-designed plants. Final report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Meliksetian, A.; Sklencar, A.M.

    1982-12-01

    The overpressure transients for Westinghouse-designed NSSSs are reviewed to determine the fluid conditions at the inlet to the PORV and safety valves. The transients considered are: licensing (FSAR) transients; extended operation of high pressure safety injection system; and cold overpressurization. The results of this review, presented in the form of tables and graphs, define the range of fluid conditions expected at the inlet to pressurized safety and power-operated relief valves utilized in Westinghouse-designed PWR units. These results will provide input to the PWR utilities in their justification that the fluid conditions under which their valve designs were tested as part of the EPRI/PWR Safety and Relief Valve Test Program indeed envelop those expected in their units.

  10. Data center cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Chainer, Timothy J; Dang, Hien P; Parida, Pritish R; Schultz, Mark D; Sharma, Arun

    2015-03-17

    A data center cooling system may include heat transfer equipment to cool a liquid coolant without vapor compression refrigeration, and the liquid coolant is used on a liquid cooled information technology equipment rack housed in the data center. The system may also include a controller-apparatus to regulate the liquid coolant flow to the liquid cooled information technology equipment rack through a range of liquid coolant flow values based upon information technology equipment temperature thresholds.

  11. Stochastic cooling in RHIC

    Energy Technology Data Exchange (ETDEWEB)

    Brennan,J.M.; Blaskiewicz, M. M.; Severino, F.

    2009-05-04

    After the success of longitudinal stochastic cooling of bunched heavy ion beam in RHIC, transverse stochastic cooling in the vertical plane of Yellow ring was installed and is being commissioned with proton beam. This report presents the status of the effort and gives an estimate, based on simulation, of the RHIC luminosity with stochastic cooling in all planes.

  12. Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

    1989-01-01

    Recent data from the HFIR vessel surveillance program indicate a substantial radiation embrittlement rate effect at low irradiation temperatures (/approximately/120/degree/F) for A212-B, A350-LF3, A105-II, and corresponding welds. PWR vessel supports are fabricated of similar materials and are subjected to the same low temperatures and fast neutron fluxes (10/sup 8/ to 10/sup 9/ neutrons/cm/sup 2//center dot/s, E > 1.0 MeV) as those in the HFIR vessel. Thus, the embrittlement rate of these structures may be greater than previously anticipated. A study sponsored by the NRC is under way at ORNL to determine the impact of the rate effect on PWR vessel-support life expectancy. The scope includes the interpretation and application of the HFIR data, a survey of all light-water-reactor vessel support designs, and a structural and fracture-mechanics analysis of the supports for two specific PWR plants of particular interest with regard to a potential for support failure as a result of propagation of flaws. Calculations performed thus far indicate best-estimate critical flaw sizes, corresponding to 32 EFPY, of /approximately/0.2 in. for one plant and /approximately/0.4 in. for the other. These flaw sizes are small enough to be of concern. However, it appears that low-cycle fatigue is not a viable mechanism for creation of flaws of this size, and thus, presumably, such flaws would have to exist at the time of fabrication. 59 refs., 128 figs., 49 tabs.

  13. Fatigue Life of Stainless Steel in PWR Environments with Strain Holding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Taesoon; Kim, Kyuhyung [KHNP CRI, Daejeon (Korea, Republic of); Seo, Myeonggyu; Jang, Changheui [KAIST, Daejeon (Korea, Republic of)

    2016-10-15

    Many components and structures of nuclear power plants are exposed to the water chemistry conditions during the operation. Recently, as design life of nuclear power plant is expanded over 60 years, the environmentally assisted fatigue (EAF) due to these water chemistry conditions has been considered as one of the important damage mechanisms of the safety class 1 components. Therefore, many studies to evaluate the effect of light water reactor (LWR) coolant environments on fatigue life of materials have been conducted. Many EAF test results including Argonne National Laboratory’s consistently indicated the substantial reduction of fatigue life in the light water reactor environments. However, there is a discrepancy between laboratory test data and plant operating experience regarding the effects of environment on fatigue: while laboratory test data suggest huge accumulation of fatigue damage, very limited experience of cracking caused by the low cycle fatigue in light water reactor. These hold-time effect tests are preformed to characterize the effects of strain holding on the fatigue life of austenitic stainless steels in PWR environments in comparison with the existing fixed strain rate results. Low cycle fatigue life tests were conducted for the type 316 stainless steel in 310℃ air and PWR environments with triangular strain. In agreement with the previous reports, the LCF life was reduced in PWR environments. Also for the slower strain rate, the reduction of LCF life was greater than the faster strain rate. The LCF test conditions for the hold-time effects were determined by the references and consideration of actual plant transient. To simulate the heat-up and cooldown transient, sub-peak strain holding during the down-hill of strain amplitude was chosen instead of peak strain holding which used in the previous researches.

  14. Application of SCALE4.4 system for burnup credit criticality analysis of PWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Hee Sung; Ro, Seung gy; Bae, Kang mok; Shin, YoungJoon; Kim, Ik Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-07-01

    An investigation on the application of burnup credit for a PWR spent fuel storage pool has been carried out with the use of the SCALE 4.4 computer code system consisting of SAS2H and CSAS6 modules in association with 44-group SCALE cross-section library. Prior to the application of the computer code system, a series of bench markings have been performed in comparison with available data. A benchmarking of the SAS2h module has been done for experimental concentration data of 54 PWR spent fuel and then correction factors with a 95% probability at a 95% confidence level have been determined on the basis of the calculated and measured concentrations of 38 nuclides. After that, the bias which might have resulted from the use of the CSAS6 module has been calculated for 46 criticality experimental data of UO{sub 2} fuel and MOX fuel assemblies. The calculation bias with one-sided tolerance limit factor (2.086) corresponding to a 95% probability at a 95% confidence level has consequently been obtained to be 0.00834. Burnup credit criticality analysis has been done for the PWR spent fuel storage pool by means of the benchmarked or validated code system. It is revealed that the minimum burnup for safe storage is 7560 MWd/tU in 5 wt% enriched fuel if both actinides and fission products in spent fuel are taken into account. However, the minimum value required seems to be 9,565 MWd/tU in the same enriched fuel provided that only the actinides are taken into consideration. (author)

  15. Evolution of reactor monitoring and protection systems for PWR; Evolution des systemes de surveillance et de protection des REP

    Energy Technology Data Exchange (ETDEWEB)

    Chaloin, B. [Electricite de France (EDF/SEPTEN), 69 - Villeurbanne (France); Mourlevat, J.L. [FRAMATOME ANP, 92 - Paris-La-Defence (France)

    2004-07-01

    This paper presents the evolution of the reactor protection systems and of the reactor monitoring systems for PWR since the initial design in the Fessenheim plant to the latest development for the EPR (European pressurized reactor). The features of both systems for the different kinds of PWR operating in France: 900 MWe, 1300 MWe and N4, are reviewed. The expected development of powerful micro-processors for computation, for data analysis and data storage will make possible in a near future the monitoring on a 3-dimensional basis and on a continuous manner, of the nuclear power released in the core. (A.C.)

  16. Use of plutonium in PWR-type reactors; Utilisation du plutonium dans les REP

    Energy Technology Data Exchange (ETDEWEB)

    Berthet, A. [Electricite de France (EDF), 75 - Paris (France). Direction de l' Equipement

    1999-04-01

    The plutonium is used, as fuel, in the pressurized water reactors. It does not exist in nature; butit is fabricated in the reactor by neutrons capture. The MOX (Mixed Oxides) is its usual name. A part is consumed by the fission, the remainder is found in the used fuel released from the reactor. The paper deals with the plutonium specificities, the research and development programs about this fuel. The technical specifications of the PWR recycling the plutonium are also included (radiation protection, reactor fueling). (A.L.B.)

  17. Research on Power Ramp Testing Method for PWR Fuel Rod at Research Reactor

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    In order to develop high performance fuel assembly for domestic nuclear power plant, it is necessary to master some fundamental test technology. So the research on the power ramp testing methods is proposed. A tentative power ramp test for short PWR fuel rod has been conducted at the heavy water research reactor (HWRR) in China Institute of Atomic Energy (CIAE) in May of 2001. The in-pile test rig was placed into the central channel of the reactor . The test rig consists of pressure pipe assembly, thimble, solid neutron absorbing screen and its driving parts, etc.. The test

  18. A study on the direct use of spent PWR fuel in CANDU reactors. DUPIC facility engineering

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Soo; Lee, Jae Sul; Choi, Jong Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This report summarizes the second year progress of phase II of DUPIC program which aims to verify experimentally the feasibility of direct use of spent PWR fuel in CANDU reactors. The project is to provide the experimental facilities and technologies that are required to perform the DUPIC experiment. As an early part of the project, engineering analysis of those facilities and construction of mock-up facility are described. Another scope of the project is to assess the DUPIC fuel cycle system and facilitate international cooperation. The progresses in this scope of work made during the fiscal year are also summarized in the report. 38 figs, 44 tabs, 8 refs. (Author).

  19. Validation of the Subchannel Code SUBCHANFLOW Using the NUPEC PWR Tests (PSBT

    Directory of Open Access Journals (Sweden)

    Uwe Imke

    2012-01-01

    Full Text Available SUBCHANFLOW is a computer code to analyze thermal-hydraulic phenomena in the core of pressurized water reactors, boiling water reactors, and innovative reactors operated with gas or liquid metal as coolant. As part of the ongoing assessment efforts, the code has been validated by using experimental data from the NUPEC PWR Subchannel and Bundle Tests (PSBT. The database includes single-phase flow bundle outlet temperature distributions, steady state and transient void distributions and critical power measurements. The performed validation work has demonstrated that the two-phase flow empirical knowledge base implemented in SUBCHANFLOW is appropriate to describe key mechanisms of the experimental investigations with acceptable accuracy.

  20. Code Development and Analysis Program: developmental checkout of the BEACON/MOD2A code. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Ramsthaler, J. A.; Lime, J. F.; Sahota, M. S.

    1978-12-01

    A best-estimate transient containment code, BEACON, is being developed by EG and G Idaho, Inc. for the Nuclear Regulatory Commission's reactor safety research program. This is an advanced, two-dimensional fluid flow code designed to predict temperatures and pressures in a dry PWR containment during a hypothetical loss-of-coolant accident. The most recent version of the code, MOD2A, is presently in the final stages of production prior to being released to the National Energy Software Center. As part of the final code checkout, seven sample problems were selected to be run with BEACON/MOD2A.

  1. Development of computational methods to describe the mechanical behavior of PWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Wanninger, Andreas; Seidl, Marcus; Macian-Juan, Rafael [Technische Univ. Muenchen, Garching (Germany). Dept. of Nuclear Engineering

    2016-10-15

    To investigate the static mechanical response of PWR fuel assemblies (FAs) in the reactor core, a structural FA model is being developed using the FEM code ANSYS Mechanical. To assess the capabilities of the model, lateral deflection tests are performed for a reference FA. For this purpose we distinguish between two environments, in-laboratory and in-reactor for different burn-ups. The results are in qualitative agreement with experimental tests and show the stiffness decrease of the FAs during irradiation in the reactor core.

  2. Cavern/Vault Disposal Concepts and Thermal Calculations for Direct Disposal of 37-PWR Size DPCs

    Energy Technology Data Exchange (ETDEWEB)

    Hardin, Ernest [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Clayton, Daniel James [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-03-01

    This report provides two sets of calculations not presented in previous reports on the technical feasibility of spent nuclear fuel (SNF) disposal directly in dual-purpose canisters (DPCs): 1) thermal calculations for reference disposal concepts using larger 37-PWR size DPC-based waste packages, and 2) analysis and thermal calculations for underground vault-type storage and eventual disposal of DPCs. The reader is referred to the earlier reports (Hardin et al. 2011, 2012, 2013; Hardin and Voegele 2013) for contextual information on DPC direct disposal alternatives.

  3. 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.

  4. Study of transient flow in fuel element of tubular plates. Accident: Shaft locking of primary cooling pump without opening the emergency gate; Estudio del regimen transitorio en el elemento combustible de placas tubulares. Accidente: Agarrotamiento de la bomba. No se abre la compuerta

    Energy Technology Data Exchange (ETDEWEB)

    Aguilas, F.; Moneva, M. A.; Garcia Ramirez, L.; Lopez Jimenez, J.; Diaz Diaz, J.

    1971-07-01

    It is analysed the thermal distribution of a fuel element of tubular plates irradiated in the JEN-1 reactor in the case of shaft locking of the primary cooling pump without opening the emergency gate. The fuel element hottest channel is studied in the position of maximum neutronic flux for three reactor power levels: 3 Hw (maximum reactor power), 2 Mw and 1 Hw. (Author) 8 refs.

  5. Laser cooling of solids

    OpenAIRE

    Nemova, Galina

    2009-01-01

    Parallel to advances in laser cooling of atoms and ions in dilute gas phase, which has progressed immensely, resulting in physics Nobel prizes in 1997 and 2001, major progress has recently been made in laser cooling of solids. I compare the physical nature of the laser cooling of atoms and ions with that of the laser cooling of solids. I point out all advantages of this new and very promising area of laser physics. Laser cooling of solids (optical refrigeration) at the present time can be lar...

  6. Evaluation of the suitabil[i]ty of Fischer-Tropsch gas-to-liquid (GTL) Primary Column Bottoms as process cooling water : analysis of microbial community dynamics, fouling, scaling and corrosion / Savia Susanna Slabbert

    OpenAIRE

    Slabbert, Savia Susanna

    2007-01-01

    Water in South Africa is becoming limiting due to economic growth, social development and the country's water demand that exceed its water availability. Water conservation in the industry can be accomplished by the reuse of process water instead of direct treatment and discharge. By reusing a process effluent as cooling water in cooling towers, the water requirements of an industry, such as Sasol, will be lower and a zero effluent discharge scenario could be achieved. At Sas...

  7. Calculation of Radioactivity and Dose Rate of Activated Corrosion Products in Water-Cooled Fusion Reactor

    Directory of Open Access Journals (Sweden)

    Jingyu Zhang

    2016-01-01

    Full Text Available In water-cooled reactor, the dominant radioactive source term under normal operation is activated corrosion products (ACPs, which have an important impact on reactor inspection and maintenance. A three-node transport model of ACPs was introduced into the new version of ACPs source term code CATE in this paper, which makes CATE capable of theoretically simulating the variation and the distribution of ACPs in a water-cooled reactor and suitable for more operating conditions. For code testing, MIT PWR coolant chemistry loop was simulated, and the calculation results from CATE are close to the experimental results from MIT, which means CATE is available and credible on ACPs analysis of water-cooled reactor. Then ACPs in the blanket cooling loop of water-cooled fusion reactor ITER under construction were analyzed using CATE and the results showed that the major contributors are the short-life nuclides, especially Mn-56. At last a point kernel integration code ARShield was coupled with CATE, and the dose rate around ITER blanket cooling loop was calculated. Results showed that after shutting down the reactor only for 8 days, the dose rate decreased nearly one order of magnitude, which was caused by the rapid decay of the short-life ACPs.

  8. The cooling conditions effects on the fraction and morphology of the epsilon primary phase in an ACuZinc5TM commercial alloy; Efecto de las condiciones de enfriamiento sobre la fraccion y morfologia de la fase primaria epsilon en una aleacion comercial ACuZinc5TM

    Energy Technology Data Exchange (ETDEWEB)

    Jareno, E. D.; Maldonado, S. I.; Hernandez, F. A.

    2012-07-01

    The mechanical properties of ACuZinc5TM alloys are influenced by the fraction solid of the primary epsilon (e) phase and their interrelations with the eta ({eta}) phase. In this study the primary epsilon phase fraction results from cooling conditions similar to that applied in different casting processes have been measured. Additionally an analysis was made from the morphology's changes and from the thermal analysis results to suggest the probably adequate moment to apply stirring in semisolid metal processing. From these results it is possible to design experimental thermal cycles to obtain an adequate distribution of e and . phases in the end castings microstructure. (Author) 21 refs.

  9. HEAT-UP AND COOL-DOWN TEMPERATURE-DEPENDENT HYDRIDE REORIENTATION BEHAVIORS IN ZIRCONIUM ALLOY CLADDING TUBES

    Directory of Open Access Journals (Sweden)

    JU-JIN WON

    2014-10-01

    Full Text Available Hydride reorientation behaviors of PWR cladding tubes under typical interim dry storage conditions were investigated with the use of as-received 250 and 485ppm hydrogen-charged Zr-Nb alloy cladding tubes. In order to evaluate the effect of typical cool-down processes on the radial hydride precipitation, two terminal heat-up temperatures of 300 and 400°C, as well as two terminal cool-down temperatures of 200 and 300°C, were considered. In addition, two cooling rates of 2.5 and 8.0°C/min during the cool-down processes were taken into account along with zero stress or a tensile hoop stress of 150MPa. It was found that the 250ppm hydrogen-charged specimen experiencing the higher terminal heat-up temperature and the lower terminal cool-down temperature generated the highest number of radial hydrides during the cool-down process under 150MPa hoop tensile stress, which may be explained by terminal solid hydrogen solubilities for precipitation, and dissolution and remaining circumferential hydrides at the terminal heat-up temperatures. In addition, the slower cool-down rate generates the larger number of radial hydrides due to a cooling rate-dependent, longer residence time at a relatively high temperature that can accelerate the radial hydride nucleation and growth.

  10. Application of the integrated analysis of safety (IAS) to sequences of Total loss of feed water in a PWR Reactor; Aplicacion del Analisis Integrado de Seguridad (ISA) a Secuencias de Perdidas Total de Agua de Alimentacion en un Reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Moreno Chamorro, P.; Gallego Diaz, C.

    2011-07-01

    The main objective of this work is to show the current status of the implementation of integrated analysis of safety (IAS) methodology and its SCAIS associated tool (system of simulation codes for IAS) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

  11. Estimate of the speed of the refrigerant on a PWR: three way based on the analysis of noise; Estimacion de la volecidad del refrigerante en un PWR: tres vias basadas en el analisis de ruido

    Energy Technology Data Exchange (ETDEWEB)

    Montalvo, C.; Ruiz, M.; Garcia Berrocal, A.

    2014-07-01

    The speed of the refrigerant is a key parameter in the monitoring of the operation a PWR. He know this value and be able to track on-site It allows an understanding of the State of the kernel with valuable information about the refrigerant, and thus behavior on heat exchange which takes place in the reactor. (Author)

  12. PWR-FBR with closed fuel cycle for a sustainable nuclear energy supply in China

    Institute of Scientific and Technical Information of China (English)

    XU Mi

    2007-01-01

    From the thermal reactor to the fast reactor and then to the fusion reactor; this is the three-step strategy that has been decided for a sustainable nuclear energy supply in China. As the main thermal reactor type, the commercialized development phase of the pressurized water reactor (PWR) has been stepped up. The development of the fast reactor (FBR) is still in the early stage, marked by China experimental fast reactor (CEFR), which is currently under construction. According to the strategy study on the fast reactor development in China, its engineering development will be divided into three steps: the CEFR with a power of 65 MWt 20 Mwe; the China prototype fast reactor (CPFR) with a power of 1 500 MWt/600 Mwe; and the China demonstration fast reactor (CDFR) with a power of 2 500-3 750 MWt 1 000-1 500 Mwe. With regards to the fuel cycle, a 100 ta PWR spent fuel reprocessing pilot plant and a 500 kg/a MOX fabrication plant are under construction. A project involving the construction of an industrial reprocessing plant and an MOX fabrication plant are also under application phase.

  13. Accelerated IGA/SCC testing of Alloy 600 in contaminated PWR environments

    Energy Technology Data Exchange (ETDEWEB)

    Miglin, B.P.; Sarver, J.M. [Babcock & Wilcox R& D Division, Alliance, OH (United States); Aoki, K. [NFI, Osaka (Japan); Koch, D.W. [Babcock & Wilcox Nuclear Services, Lynchburg, VA (United States); Takamatsu, H. [Kansai Electric, Osaka (Japan)

    1992-12-31

    An accelerated corrosion test (360{degrees}C for 2000 hrs) was performed on C-ring specimens machined from one heat of Alloy 600 tubing in the mill-annealed condition. The specimens were exposed to secondary-side pressurized-water-reactor (PWR) solutions contaminated with lead, sulfur, silicon, and a combination of these contaminants. Where possible, MULTEQ calculations were performed to determine the chemical concentrations so that a constant elevated-temperature pH of 4.5 was achieved. This test was designed to examine the ability of these contaminants to cause intergranular attack and/or stress corrosion in stressed Alloy 600 tubing. The results from this test demonstrated that under the test conditions used, lead-contaminated PWR secondary water induces and propagates intergranular attack (IGA) and stress corrosion cracking (SCC) in Alloy 600. Attack was intergranular; the degree of attack did not vary in the liquid or vapor portions of the test environments. Although attack was more severe at higher stresses, significant attack was observed in samples stressed to the typical operating stress. Solutions of only sulfur and only silicon displayed no initiation or propagation of either IGA or SCC. However, the solution containing all three contaminants caused attack with identical morphology to that observed in the lead-contaminated solution.

  14. Vulnerability of a partially flooded PWR reactor cavity to a steam explosion

    Energy Technology Data Exchange (ETDEWEB)

    Cizelj, Leon [' Jozef Stefan' Institute Jamova 39, SI 1000 Ljubljana (Slovenia)]. E-mail: leon.cizelj@ijs.si; Koncar, Bostjan [' Jozef Stefan' Institute Jamova 39, SI 1000 Ljubljana (Slovenia); Leskovar, Matjaz [' Jozef Stefan' Institute Jamova 39, SI 1000 Ljubljana (Slovenia)

    2006-08-15

    When the hot molten core comes into contact with the water in the reactor cavity a steam explosion may occur. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and later, during the expansion of the water vapour, to production of missiles that may endanger surrounding structures. The purpose of the performed analysis is to provide an estimation of the expected pressure loadings on the typical PWR cavity structures during a steam explosion, and to make an assessment of the vulnerabilities of the typical PWR cavity structures to steam explosions. To achieve this, the fit-for-purpose steam explosion model is proposed, followed by comprehensive and reasonably conservative analyses of two typical ex-vessel steam explosion cases differing in the steam explosion energy conversion ratio. In particular, the vulnerability of the surrounding reinforced concrete walls to damage has been sought in both cases.

  15. Analysis of bubble pressure in the rim region of high burnup PWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-02-01

    Bubble pressure in the rim region of high burnup PWR UO{sub 2} fuel has been modeled based on measured rim width, porosity and bubble density. Using the assumption that excessive bubble pressure in the rim is inversely proportional to its radius, proportionality constant is derived as a function of average pellet burnup and bubble radius. This approach is possible because the integration of the number of Xe atoms retained in the rim bubbles, which can be calculated as a function of bubble radius, over the bubble radius gives the total number of Xe atoms in the rim bubbles. Here the total number of Xe atoms in the rim bubbles can be derived from the measured Xe depletion fraction in the matrix and the calculated rim thickness. Then the rim bubble pressure is obtained as a function of fuel burnup and bubble size from the proportionality constant. Therefore, the present model can provide some useful information that would be required to analyze the behavior of high burnup PWR UO{sub 2} fuel under both normal and transient operating conditions. 28 refs., 9 figs. (Author)

  16. Construction and utilization of linear empirical core models for PWR in-core fuel management

    Energy Technology Data Exchange (ETDEWEB)

    Okafor, K.C.

    1988-01-01

    An empirical core-model construction procedure for pressurized water reactor (PWR) in-core fuel management is developed that allows determining the optimal BOC k{sub {infinity}} profiles in PWRs as a single linear-programming problem and thus facilitates the overall optimization process for in-core fuel management due to algorithmic simplification and reduction in computation time. The optimal profile is defined as one that maximizes cycle burnup. The model construction scheme treats the fuel-assembly power fractions, burnup, and leakage as state variables and BOC zone enrichments as control variables. The core model consists of linear correlations between the state and control variables that describe fuel-assembly behavior in time and space. These correlations are obtained through time-dependent two-dimensional core simulations. The core model incorporates the effects of composition changes in all the enrichment control zones on a given fuel assembly and is valid at all times during the cycle for a given range of control variables. No assumption is made on the geometry of the control zones. A scatter-composition distribution, as well as annular, can be considered for model construction. The application of the methodology to a typical PWR core indicates good agreement between the model and exact simulation results.

  17. Control rod ejection accident analysis for a PWR with thorium fuel loading

    Energy Technology Data Exchange (ETDEWEB)

    Da Cruz, D.F. [Nuclear Research and Consultancy Group NRG, Westerduinweg 3, P.O. Box 25, 1755 ZG Petten (Netherlands)

    2010-07-01

    This paper presents the results of 3-D transient analysis of a pressurized water reactor (PWR) core loaded with 100% Th-Pu MOX fuel assemblies. The aim of this study is to evaluate the safety impact of applying a full loading of this innovative fuel in PWRs of the current generation. A reactivity insertion accident scenario has been simulated using the reactor core analysis code PANTHER, used in conjunction with the lattice code WIMS. A single control rod assembly, with the highest reactivity worth, has been considered to be ejected from the core within 100 milliseconds, which may occur due to failure of the casing of the control rod driver mechanism. Analysis at both hot full power and hot zero power reactor states have been taken into account. The results were compared with those obtained for a representative PWR fuelled with UO{sub 2} fuel assemblies. In general the results obtained for both cores were comparable, with some differences associated mainly to the harder neutron spectrum observed for the Th-Pu MOX core, and to some specific core design features. The study has been performed as part of the LWR-DEPUTY project of the EURATOM 6. Framework Programme, where several aspects of novel fuels are being investigated for deep burning of plutonium in existing nuclear power plants. (authors)

  18. NODAL3 Sensitivity Analysis for NEACRP 3D LWR Core Transient Benchmark (PWR

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2016-01-01

    Full Text Available This paper reports the results of sensitivity analysis of the multidimension, multigroup neutron diffusion NODAL3 code for the NEACRP 3D LWR core transient benchmarks (PWR. The code input parameters covered in the sensitivity analysis are the radial and axial node sizes (the number of radial node per fuel assembly and the number of axial layers, heat conduction node size in the fuel pellet and cladding, and the maximum time step. The output parameters considered in this analysis followed the above-mentioned core transient benchmarks, that is, power peak, time of power peak, power, averaged Doppler temperature, maximum fuel centerline temperature, and coolant outlet temperature at the end of simulation (5 s. The sensitivity analysis results showed that the radial node size and maximum time step give a significant effect on the transient parameters, especially the time of power peak, for the HZP and HFP conditions. The number of ring divisions for fuel pellet and cladding gives negligible effect on the transient solutions. For productive work of the PWR transient analysis, based on the present sensitivity analysis results, we recommend NODAL3 users to use 2×2 radial nodes per assembly, 1×18 axial layers per assembly, the maximum time step of 10 ms, and 9 and 1 ring divisions for fuel pellet and cladding, respectively.

  19. An extension of the validation of SCALE (SAS2H) isotopic predictions for PWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.; Hermann, O.W.

    1996-09-01

    Isotopic characterization of spent fuel via depletion and decay calculations is necessary for determination of source terms. Unlike fresh fuel assumptions typically used in criticality safety analysis of spent fuel configurations, burnup credit applications also rely on depletion and decay calculations to predict spent fuel composition; these isotopics are used in subsequent criticality calculations to assess the reduced worth of spent fuel. To validate the depletion codes and data, experiment is compared with predictions; such comparisons have been done in earlier ORNL work. This report describes additional independent measurements and corresponding calculations as a supplement. The current work includes measured isotopic data from 19 spent fuel samples from the Italian Trino Vercelles PWR and the US Turkey Point-3 PWR. In addition, an approach to determine biases and uncertainties between calculated and measured isotopic concentrations is discussed, together with a method to statistically combine these terms to obtain a conservative estimate of spent fuel isotopic concentrations. Results on combination of measured-to-calculated ratios are presented. The results described herein represent an extension to a new reactor design and spent fuel samples with enrichment as high as 3.9 wt% {sup 235}U. Consistency with the earlier work for each of two different cross-section libraries suggests that the estimated biases for each of the isotopes in the earlier work are reasonably good estimates.

  20. On applicability of plate and shell heat exchangers for steam generation in naval PWR

    Energy Technology Data Exchange (ETDEWEB)

    Freire, Luciano Ondir, E-mail: luciano.ondir@gmail.com; Andrade, Delvonei Alves de, E-mail: delvonei@ipen.br

    2014-12-15

    Highlights: • Given emissions restrictions, nuclear propulsion may be an alternative. • Plate and shell heat exchangers (PSHE) are a mature technology on market. • PSHE are compact and could be used as steam generators. • Preliminary calculations to obtain a PWR for a large container ship are performed. • Results suggest PSHE improve overall compactness and cost. - Abstract: The pressure on reduction of gas emissions is going to raise the price of fossil fuels and an alternative to fossil fuels is nuclear energy. Naval reactors have some differences from stationary PWR because they have limitations on volume and weight, requiring compact solutions. On the other hand, a source of problems in naval reactors across history is the steam generation function. In order to reduce nuclear containment footprint, it is desirable to employ integral designs, which, however, poses complications and design constraints for recirculation type steam generators, being interesting to employ once through steam generators, whose historic at Babcock and Wilcox is better than recirculation steam generators. Plate and shell heat exchangers are a mature technology made available by many suppliers which allows heat exchange at high temperature and pressure. This work investigates the feasibility of the use of an array of welded plate heat exchangers of a material approved by ASME for pressure barrier (Ti-3Al-2.5V) in a hypothetical naval reactor. It was found it is feasible from thermal-hydraulic point of view and presents advantages over other steam generator designs.

  1. Gas entrainment by one single French PWR spray, SARNET-2 spray benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Malet, J., E-mail: jeanne.malet@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, Saclay (France); Mimouni, S., E-mail: stephane.mimouni@edf.fr [Electricité de France, EDF MF2E, Chatou (France); Manzini, G., E-mail: giovanni.manzini@rse-web.it [RSE, Milano (Italy); Xiao, J., E-mail: jianjun.xiao@kit.edu [IKET, KIT, Karlsruhe (Germany); Vyskocil, L., E-mail: vyl@ujv.cz [UJV Rez (Czech Republic); Siccama, N.B., E-mail: siccama@nrg.eu [NRG, Safety and Power (Netherlands); Huhtanen, R., E-mail: risto.huhtanen@vtt.fi [VTT, PO Box 1000, FI-02044 VTT (Finland)

    2015-02-15

    Highlights: • This paper presents a benchmark performed in the frame of the SARNET-2 EU project. • It concerns momentum transfer between a PWR spray and the surrounding gas. • The entrained gas velocities can vary up to 100% from one code to another. • Simplified boundary conditions for sprays are generally used by the code users. • It is shown how these simplified conditions impact the gas entrainment. - Abstract: This paper presents a benchmark performed in the frame of the SARNET-2 EU project, dealing with momentum transfer between a real-scale PWR spray and the surrounding gas. It presents a description of the IRSN tests on the CALIST facility, the participating codes (8 contributions), code-experiment and code-to-code comparisons. It is found that droplet velocities are almost well calculated one meter below the spray nozzle, even if the spread of the spray is not recovered and the values of the entrained gas velocity vary up to 100% from one code to another. Concerning sensitivity analysis, several ‘simplifications’ have been made by the contributors, especially based on the boundary conditions applied at the location where droplets are injected. It is shown here that such simplifications influence droplet and entrained gas characteristics. The next step will be to translate these conclusions in terms of variables representative of interesting parameters for nuclear safety.

  2. Stochastic cooling in RHIC

    Energy Technology Data Exchange (ETDEWEB)

    Brennan J. M.; Blaskiewicz, M.; Mernick, K.

    2012-05-20

    The full 6-dimensional [x,x'; y,y'; z,z'] stochastic cooling system for RHIC was completed and operational for the FY12 Uranium-Uranium collider run. Cooling enhances the integrated luminosity of the Uranium collisions by a factor of 5, primarily by reducing the transverse emittances but also by cooling in the longitudinal plane to preserve the bunch length. The components have been deployed incrementally over the past several runs, beginning with longitudinal cooling, then cooling in the vertical planes but multiplexed between the Yellow and Blue rings, next cooling both rings simultaneously in vertical (the horizontal plane was cooled by betatron coupling), and now simultaneous horizontal cooling has been commissioned. The system operated between 5 and 9 GHz and with 3 x 10{sup 8} Uranium ions per bunch and produces a cooling half-time of approximately 20 minutes. The ultimate emittance is determined by the balance between cooling and emittance growth from Intra-Beam Scattering. Specific details of the apparatus and mathematical techniques for calculating its performance have been published elsewhere. Here we report on: the method of operation, results with beam, and comparison of results to simulations.

  3. Stress corrosion cracking in the vessel closure head penetrations of French PWR`s; Fissuration par corrosion sous contrainte de penetrations de couvercle de cuve de reacteur nucleaire francais a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Buisine, D.; Cattant, F.; Champredonde, J.; Pichon, C.; Benhamou, C.; Gelpi, A.; Vaindirlis, M.

    1994-01-01

    During a hydrotest in September 1991, part of the statutory decennial in-service inspection, a leak was detected on the vessel head of Bugey 3, which is one of the first 900 MW 3-loop PWR`s in France. This leak was due to a cracked penetration used for a control rod drive mechanism. The investigations performed identified Primary Stress Corrosion Cracking of Alloy 600 as being the origin of this degradation. So a lot of the same design PWR`s are a concern due to this generic problem. In this case, PWSCC was linked to: - hot temperature of the vessel head; - high residual stresses due to the welding process between peripherical penetrations and the vessel head; - sensitivity of forged Alloy 600 used for penetration manufacturing. This following paper will present the cracked analysis based, in particular, on the main results obtained in France on each of these items. These results come from the operating experience, the destructive examinations and the programs which are running on stress analysis and metallurgical characterizations. (authors). 9 figs., 2 tabs.

  4. Effect of sensitization and cold work on stress corrosion susceptibility of austenitic stainless steels in BWR and PWR conditions

    Energy Technology Data Exchange (ETDEWEB)

    Haenninen, H.; Aho-Mantila, I.

    1981-05-01

    The influence of metallurgical variables on stress corrosion cracking of austenitic stainless steels, in particular AISI 304 and OX18H10T, has been examined both in O2-enriched BWR-conditions (8 ppm O2) and in typical PWR-conditions.

  5. Cooling by Thermodynamic Induction

    Science.gov (United States)

    Patitsas, S. N.

    2017-03-01

    A method is described for cooling conductive channels to below ambient temperature. The thermodynamic induction principle dictates that the electrically biased channel will cool if the electrical conductance decreases with temperature. The extent of this cooling is calculated in detail for both cases of ballistic and conventional transport with specific calculations for carbon nanotubes and conventional metals, followed by discussions for semiconductors, graphene, and metal-insulator transition systems. A theorem is established for ballistic transport stating that net cooling is not possible. For conventional transport, net cooling is possible over a broad temperature range, with the range being size-dependent. A temperature clamping scheme for establishing a metastable nonequilibrium stationary state is detailed and followed with discussion of possible applications to on-chip thermoelectric cooling in integrated circuitry and quantum computer systems.

  6. Cooling by Thermodynamic Induction

    Science.gov (United States)

    Patitsas, S. N.

    2016-11-01

    A method is described for cooling conductive channels to below ambient temperature. The thermodynamic induction principle dictates that the electrically biased channel will cool if the electrical conductance decreases with temperature. The extent of this cooling is calculated in detail for both cases of ballistic and conventional transport with specific calculations for carbon nanotubes and conventional metals, followed by discussions for semiconductors, graphene, and metal-insulator transition systems. A theorem is established for ballistic transport stating that net cooling is not possible. For conventional transport, net cooling is possible over a broad temperature range, with the range being size-dependent. A temperature clamping scheme for establishing a metastable nonequilibrium stationary state is detailed and followed with discussion of possible applications to on-chip thermoelectric cooling in integrated circuitry and quantum computer systems.

  7. Radiant Floor Cooling Systems

    DEFF Research Database (Denmark)

    Olesen, Bjarne W.

    2008-01-01

    In many countries, hydronic radiant floor systems are widely used for heating all types of buildings such as residential, churches, gymnasiums, hospitals, hangars, storage buildings, industrial buildings, and smaller offices. However, few systems are used for cooling.This article describes a floor...... cooling system that includes such considerations as thermal comfort of the occupants, which design parameters will influence the cooling capacity and how the system should be controlled. Examples of applications are presented....

  8. Initial Cooling Experiment (ICE)

    CERN Multimedia

    Photographic Service

    1978-01-01

    In 1977, in a record-time of 9 months, the magnets of the g-2 experiment were modified and used to build a proton/antiproton storage ring: the "Initial Cooling Experiment" (ICE). It served for the verification of the cooling methods to be used for the "Antiproton Project". Stochastic cooling was proven the same year, electron cooling followed later. Also, with ICE the experimental lower limit for the antiproton lifetime was raised by 9 orders of magnitude: from 2 microseconds to 32 hours. For its previous life as g-2 storage ring, see 7405430. More on ICE: 7711282, 7809081, 7908242.

  9. High energy electron cooling

    Energy Technology Data Exchange (ETDEWEB)

    Parkhomchuk, V. [Budker Institute of Nuclear Physics, Novosibirsk (Russian Federation)

    1997-09-01

    High energy electron cooling requires a very cold electron beam. The questions of using electron cooling with and without a magnetic field are presented for discussion at this workshop. The electron cooling method was suggested by G. Budker in the middle sixties. The original idea of the electron cooling was published in 1966. The design activities for the NAP-M project was started in November 1971 and the first run using a proton beam occurred in September 1973. The first experiment with both electron and proton beams was started in May 1974. In this experiment good result was achieved very close to theoretical prediction for a usual two component plasma heat exchange.

  10. Power electronics cooling apparatus

    Science.gov (United States)

    Sanger, Philip Albert; Lindberg, Frank A.; Garcen, Walter

    2000-01-01

    A semiconductor cooling arrangement wherein a semiconductor is affixed to a thermally and electrically conducting carrier such as by brazing. The coefficient of thermal expansion of the semiconductor and carrier are closely matched to one another so that during operation they will not be overstressed mechanically due to thermal cycling. Electrical connection is made to the semiconductor and carrier, and a porous metal heat exchanger is thermally connected to the carrier. The heat exchanger is positioned within an electrically insulating cooling assembly having cooling oil flowing therethrough. The arrangement is particularly well adapted for the cooling of high power switching elements in a power bridge.

  11. Conceptual Design of Structural Components of a Dual Cooled Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung-Kyu; Lee, Young-Ho; Lee, Kang-Hee; Kim, Jae-Yong; Yoon, Kyung-Ho

    2008-01-15

    A dual cooled fuel, featured by an internal as well as an external coolant flow passage of a fuel rod, was suggested to enable a large-scaled power-uprate of PWR plant and launched as one of the National Nuclear R and D Projects in 2007. It is necessary to make the dual cooled fuel be compatible with an OPR-1000 system to maximize the economy. Also, the structural components of the dual cooled fuel should be designed to realize their features. To this end, a conceptual design of a spacer grid, outer and center guide tubes, and top and bottom end pieces has been carried out in the project 'Development of Design Technology for Dual Cooled Fuel Structure'. For the spacer grids, it is suggested that springs and dimples are located at or near the cross points of the straps due to a considerably narrowed rod-to-rod gap. Candidate shapes of the grids were also developed and applied for domestic patents. For the outer and center guide tubes, a dual tube like a fuel rod was suggested to make the subchannel areas around the guide tubes be similar to those around the fuel rods of enlarged diameter. It was applied for the domestic patent as well. For the top and bottom end pieces, the shape and pattern have been changed from the conventional ones reflecting the fuel rods' changes. Technical issues and method of resolution for each components were listed up for a basic design works in the following years.

  12. Cooling of electronics in collider experiments

    Energy Technology Data Exchange (ETDEWEB)

    Richard P. Stanek et al.

    2003-11-07

    Proper cooling of detector electronics is critical to the successful operation of high-energy physics experiments. Collider experiments offer unique challenges based on their physical layouts and hermetic design. Cooling systems can be categorized by the type of detector with which they are associated, their primary mode of heat transfer, the choice of active cooling fluid, their heat removal capacity and the minimum temperature required. One of the more critical detector subsystems to require cooling is the silicon vertex detector, either pixel or strip sensors. A general design philosophy is presented along with a review of the important steps to include in the design process. Factors affecting the detector and cooling system design are categorized. A brief review of some existing and proposed cooling systems for silicon detectors is presented to help set the scale for the range of system designs. Fermilab operates two collider experiments, CDF & D0, both of which have silicon systems embedded in their detectors. A review of the existing silicon cooling system designs and operating experience is presented along with a list of lessons learned.

  13. Stochastic cooling of a high energy collider

    Energy Technology Data Exchange (ETDEWEB)

    Blaskiewicz, M.; Brennan, J.M.; Lee, R.C.; Mernick, K.

    2011-09-04

    Gold beams in RHIC revolve more than a billion times over the course of a data acquisition session or store. During operations with these heavy ions the event rates in the detectors decay as the beams diffuse. A primary cause for this beam diffusion is small angle Coloumb scattering of the particles within the bunches. This intra-beam scattering (IBS) is particularly problematic at high energy because the negative mass effect removes the possibility of even approximate thermal equilibrium. Stochastic cooling can combat IBS. A theory of bunched beam cooling was developed in the early eighties and stochastic cooling systems for the SPS and the Tevatron were explored. Cooling for heavy ions in RHIC was also considered.

  14. Elastocaloric cooling: Stretch to actively cool

    Science.gov (United States)

    Ossmer, Hinnerk; Kohl, Manfred

    2016-10-01

    The elastocaloric effect can be exploited in solid-state cooling technologies as an alternative to conventional vapour compression. Now, an elastocaloric device based on the concept of active regeneration achieves a temperature lift of 15.3 K and efficiencies competitive with other caloric-based approaches.

  15. Measure Guideline: Ventilation Cooling

    Energy Technology Data Exchange (ETDEWEB)

    Springer, D.; Dakin, B.; German, A.

    2012-04-01

    The purpose of this measure guideline on ventilation cooling is to provide information on a cost-effective solution for reducing cooling system energy and demand in homes located in hot-dry and cold-dry climates. This guideline provides a prescriptive approach that outlines qualification criteria, selection considerations, and design and installation procedures.

  16. The final cool down

    CERN Multimedia

    Thursday 29th May, the cool-down of the final sector (sector 4-5) of LHC has begun, one week after the start of the cool-down of sector 1-2. It will take five weeks for the sectors to be cooled from room temperature to 5 K and a further two weeks to complete the cool down to 1.9 K and the commissioning of cryogenic instrumentation, as well as to fine tune the cryogenic plants and the cooling loops of cryostats.Nearly a year and half has passed since sector 7-8 was cooled for the first time in January 2007. For Laurent Tavian, AT/CRG Group Leader, reaching the final phase of the cool down is an important milestone, confirming the basic design of the cryogenic system and the ability to operate complete sectors. “All the sectors have to operate at the same time otherwise we cannot inject the beam into the machine. The stability and reliability of the cryogenic system and its utilities are now very important. That will be the new challenge for the coming months,” he explains. The status of the cool down of ...

  17. Solar absorption cooling

    NARCIS (Netherlands)

    Kim, D.-S.

    2007-01-01

    As the world concerns more and more on global climate changes and depleting energy resources, solar cooling technology receives increasing interests from the public as an environment-friendly and sustainable alternative. However, making a competitive solar cooling machine for the market still

  18. Passive evaporative cooling

    NARCIS (Netherlands)

    Tzoulis, A.

    2011-01-01

    This "designers' manual" is made during the TIDO-course AR0531 Smart & Bioclimatic Design. Passive techniques for cooling are a great way to cope with the energy problem of the present day. This manual introduces passive cooling by evaporation. These methods have been used for many years in traditi

  19. Data center cooling method

    Energy Technology Data Exchange (ETDEWEB)

    Chainer, Timothy J.; Dang, Hien P.; Parida, Pritish R.; Schultz, Mark D.; Sharma, Arun

    2015-08-11

    A method aspect for removing heat from a data center may use liquid coolant cooled without vapor compression refrigeration on a liquid cooled information technology equipment rack. The method may also include regulating liquid coolant flow to the data center through a range of liquid coolant flow values with a controller-apparatus based upon information technology equipment temperature threshold of the data center.

  20. Liquid Cooled Garments

    Science.gov (United States)

    1979-01-01

    Astronauts working on the surface of the moon had to wear liquid-cooled garments under their space suits as protection from lunar temperatures which sometimes reach 250 degrees Fahrenheit. In community service projects conducted by NASA's Ames Research Center, the technology developed for astronaut needs has been adapted to portable cooling systems which will permit two youngsters to lead more normal lives.

  1. Solar absorption cooling

    NARCIS (Netherlands)

    Kim, D.-S.

    2007-01-01

    As the world concerns more and more on global climate changes and depleting energy resources, solar cooling technology receives increasing interests from the public as an environment-friendly and sustainable alternative. However, making a competitive solar cooling machine for the market still remain

  2. Coherent electron cooling

    Energy Technology Data Exchange (ETDEWEB)

    Litvinenko,V.

    2009-05-04

    Cooling intense high-energy hadron beams remains a major challenge in modern accelerator physics. Synchrotron radiation is still too feeble, while the efficiency of two other cooling methods, stochastic and electron, falls rapidly either at high bunch intensities (i.e. stochastic of protons) or at high energies (e-cooling). In this talk a specific scheme of a unique cooling technique, Coherent Electron Cooling, will be discussed. The idea of coherent electron cooling using electron beam instabilities was suggested by Derbenev in the early 1980s, but the scheme presented in this talk, with cooling times under an hour for 7 TeV protons in the LHC, would be possible only with present-day accelerator technology. This talk will discuss the principles and the main limitations of the Coherent Electron Cooling process. The talk will describe the main system components, based on a high-gain free electron laser driven by an energy recovery linac, and will present some numerical examples for ions and protons in RHIC and the LHC and for electron-hadron options for these colliders. BNL plans a demonstration of the idea in the near future.

  3. Modeling gasodynamic vortex cooling

    Science.gov (United States)

    Allahverdyan, A. E.; Fauve, S.

    2017-08-01

    We aim at studying gasodynamic vortex cooling in an analytically solvable, thermodynamically consistent model that can explain limitations on the cooling efficiency. To this end, we study an angular plus radial flow between two (coaxial) rotating permeable cylinders. Full account is taken of compressibility, viscosity, and heat conductivity. For a weak inward radial flow the model qualitatively describes the vortex cooling effect, in terms of both temperature and the decrease of the stagnation enthalpy, seen in short uniflow vortex (Ranque) tubes. The cooling does not result from external work and its efficiency is defined as the ratio of the lowest temperature reached adiabatically (for the given pressure gradient) to the lowest temperature actually reached. We show that for the vortex cooling the efficiency is strictly smaller than 1, but in another configuration with an outward radial flow, we find that the efficiency can be larger than 1. This is related to both the geometry and the finite heat conductivity.

  4. Hydronic rooftop cooling systems

    Science.gov (United States)

    Bourne, Richard C.; Lee, Brian Eric; Berman, Mark J.

    2008-01-29

    A roof top cooling unit has an evaporative cooling section that includes at least one evaporative module that pre-cools ventilation air and water; a condenser; a water reservoir and pump that captures and re-circulates water within the evaporative modules; a fan that exhausts air from the building and the evaporative modules and systems that refill and drain the water reservoir. The cooling unit also has a refrigerant section that includes a compressor, an expansion device, evaporator and condenser heat exchangers, and connecting refrigerant piping. Supply air components include a blower, an air filter, a cooling and/or heating coil to condition air for supply to the building, and optional dampers that, in designs that supply less than 100% outdoor air to the building, control the mixture of return and ventilation air.

  5. INITIAL COOLING EXPERIMENT (ICE)

    CERN Multimedia

    1979-01-01

    ICE was built in 1977, using the modified bending magnets of the g-2 muon storage ring (see 7405430). Its purpose was to verify the validity of stochastic and electron cooling for the antiproton project. Stochastic cooling proved a resounding success early in 1978 and the antiproton project could go ahead, now entirely based on stochastic cooling. Electron cooling was experimented with in 1979. The 26 kV equipment is housed in the cage to the left of the picture, adjacent to the "e-cooler" located in a straight section of the ring. With some modifications, the cooler was later transplanted into LEAR (Low Energy Antiproton Ring) and then, with further modifications, into the AD (Antiproton Decelerator), where it cools antiprotons to this day (2006). See also: 7711282, 7802099, 7809081.

  6. INITIAL COOLING EXPERIMENT (ICE)

    CERN Multimedia

    1978-01-01

    ICE was built in 1977, in a record time of 9 months, using the modified bending magnets of the g-2 muon storage ring. Its purpose was to verify the validity of stochastic and electron cooling for the antiproton project, to be launched in 1978. Already early in 1978, stochastic cooling proved a resounding success, such that the antiproton (p-pbar)project was entirely based on it. Tests of electron cooling followed later: protons of 46 MeV kinetic energy were cooled with an electron beam of 26 kV and 1.3 A. The cage seen prominently in the foreground houses the HV equipment, adjacent to the "cooler" installed in a straight section of the ring. With some modifications, the cooler was later transplanted into LEAR (Low Energy Antiproton Ring) and then, with further modifications, into the AD (Antiproton Decelerator), where it cools antiprotons to this day (2006). See also: 7711282, 7802099, 7908242.

  7. 改进型快谱超临界水冷堆增殖特性初步研究%Primary Study on Breeding Property of Improved Supercritical Water Cooled Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    刘紫静; 于涛; 谢金森

    2012-01-01

    In this paper, the core mode of improved supercritical water cooled fast reactor is established. At first, reasonable fuel assembly design is obtained by studying the influences of seed fuel pin diameter and blanket coolant channel diameter to conversion ratio (Cr). Then, viod reactivity coefficient and CR of six different core arrangements are calculated. Finaly, the influences of fuel components to CR and void reactivity coefficient are analysed. The results show that negative void reactivity coefficient can be satisfied and Cr can be increased by reducing Hydrogen to Heavy-metal ratio (H/HM), increasing blanket assembly numbers by proper distribution. Cr is substantially increased and more negative void reactivity coefficient can be met by reducing PuO2 mass ratio in fuel, when PuO2 mass ratio reach 20.8% in MOX fuel and 235U enriched at 0.2% in UO2 fuel have been adopted as seed and blanket assmbly respectively, the sixth core program reaches CR=1.04395 and give negative void reactivity coefficient, which meets the primary requirements for SCFR breeding.%建立改进型快谱超临界水冷堆( SCFR-M)堆芯模型,探讨点火区燃料棒直径和增殖区水棒直径对堆芯转换比的影响,得到合理的燃料组件设计形式.设计计算6种不同堆芯布置下的增殖特性和空泡反应性,分析燃料组分对堆芯转换比和空泡反应性系数的影响.结果表明:减小氢原子数与重金属原子数之比(H/HM),增加堆芯增殖组件数目并采用合理布置可在满足负空泡反应系数的同时提高转换比;降低燃料中PuO2质量分数可以使转换比大幅增加,同时使堆芯的空泡反应性系数有更大负值;当点火组件采用PuO2质量分数为20.8%的MOX燃料,增殖组件采用235U富集度为0.2%的UO2燃料,方案6的设计可以使堆芯的初始转换比达到1.04395,并且空泡反应性系数为负,初步达到快谱超临界水冷堆的增殖要求.

  8. Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Aldrich, D.C.; McGrath, P.E.; Rasmussen, N.C.

    1978-06-01

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment failure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects.

  9. Effect of Weld Properties on the Crush Strength of the PWR Spacer Grid

    Directory of Open Access Journals (Sweden)

    Kee-nam Song

    2012-01-01

    Full Text Available Mechanical properties in a weld zone are different from those in the base material because of different microstructures. A spacer grid in PWR fuel is a structural component with an interconnected and welded array of slotted grid straps. Previous research on the strength analyses of the spacer grid was performed using base material properties owing to a lack of mechanical properties in the weld zone. In this study, based on the mechanical properties in the weld zone of the spacer grid recently obtained by an instrumented indentation technique, the strength analyses considering the mechanical properties in the weld zone were performed, and the analysis results were compared with the previous research.

  10. Sensitivity analysis of a PWR fuel element using zircaloy and silicon carbide claddings

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Rochkhudson B. de; Cardoso, Fabiano; Salome, Jean A.D.; Pereira, Claubia; Fortini, Angela, E-mail: rochkhudson@ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Escola de Engenharia. Departamento de Engenharia Nuclear

    2015-07-01

    The alloy composed of zirconium has been used effectively for over 50 years in claddings of nuclear fuel, especially for PWR type reactors. However, to increase fuel enrichment with the aim of raising the burning and maintaining the safety of nuclear plants is of great relevance the study of new materials that can replace safely and efficiently zircaloy cladding. Among several proposed material, silicon carbide (SiC) has a potential to replace zircaloy as fuel cladding material due to its high-temperature tolerance, chemical stability and low neutron affinity. In this paper, the goal is to expand the study with silicon carbide cladding, checking its behavior when submitted to an environment with boron, burnable poison rods, and temperature variations. Sensitivity calculation and the impact in multiplication factor to both claddings, zircaloy and silicon carbide, were performed during the burnup. The neutronic analysis was made using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation) code. (author)

  11. San Onofre PWR Data for Code Validation of MOX Fuel Depletion Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.

    1999-09-01

    The isotopic composition of mixed-oxide fuel (fabricated with both uranium and plutonium isotope) discharged from reactors is of interest to the Fissile Material Disposition Program. The validation of depletion codes used to predict isotopic compositions of MOX fuel, similar to studies concerning uranium-only fueled reactors, thus, is very important. The EEI-Westinghouse Plutonium Recycle Demonstration Program was conducted to examine the use of MOX fuel in the San Onofre PWR, Unit I, during cycles 2 and 3. The data usually required as input to depletion codes, either one-dimensional or lattice codes, were taken from various sources and compiled into this report. Where data were either lacking or determined inadequate, the appropriate data were supplied from other references. The scope of the reactor operations and design data, in addition to the isotopic analyses, were considered to be of sufficient quality for depletion code validation.

  12. Steady characteristic investigation on passive residual heat removal system of Chinese advanced PWR

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    Thermal-hydraulic characteristic investigation on passive residual heat removal system(PRHRS)of Chinese advanced PWR was conducted to provide input data for PRHRS design and to demonstrate the feasibility of unique design features.A total of 237 sets of test data at steady state have been obtained and the main influence factors on the two-phase natural circulation flow rate and residual heat removal capability were identified.On the basis of theory analysis,a correlation of two-phase natural circulation was obtained,and relative errors of 95% test data were less than±16%.There is a considerable effect of the system status parameters on the threshold of height between heat source and heat sink,and its correlation of two-phase natural circulation system has been obtained.The steady characteristic research shows that PRHRS has the capability of removing the core decay power through natural circulation.

  13. Improving PWR core simulations by Monte Carlo uncertainty analysis and Bayesian inference

    CERN Document Server

    Castro, Emilio; Buss, Oliver; Garcia-Herranz, Nuria; Hoefer, Axel; Porsch, Dieter

    2016-01-01

    A Monte Carlo-based Bayesian inference model is applied to the prediction of reactor operation parameters of a PWR nuclear power plant. In this non-perturbative framework, high-dimensional covariance information describing the uncertainty of microscopic nuclear data is combined with measured reactor operation data in order to provide statistically sound, well founded uncertainty estimates of integral parameters, such as the boron letdown curve and the burnup-dependent reactor power distribution. The performance of this methodology is assessed in a blind test approach, where we use measurements of a given reactor cycle to improve the prediction of the subsequent cycle. As it turns out, the resulting improvement of the prediction quality is impressive. In particular, the prediction uncertainty of the boron letdown curve, which is of utmost importance for the planning of the reactor cycle length, can be reduced by one order of magnitude by including the boron concentration measurement information of the previous...

  14. A comparison of the CHF between tubes and annuli under PWR thermal-hydraulic conditions

    Energy Technology Data Exchange (ETDEWEB)

    Herer, C. [RRAMATOME EP/TC, Paris (France); Souyri, A. [EdF DER/RNE/TTA, Chatou (France); Garnier, J. [CEA DRN/DTP/STR/LETC, Grenoble (France)

    1995-09-01

    Critical Heat Flux (CHF) tests were carried out in three tubes with inside diameters of 8, 13, and 19.2 mm and in two annuli with an inner tube of 9.5 mm and an outer tube of 13 or 19.2 mm. All axial heat flux distributions in the test sections were uniform. The coolant fluid was Refrigerant 12 (Freon-12) under PWR thermal-hydraulic conditions (equivalent water conditions - Pressure: 7 to 20 MPa, Mass Velocity: 1000 to 6000 kg/m2/s, Local Quality: -75% to +45%). The effect of tube diameter is correlated for qualities under 15%. The change from the tube to the annulus configuration is correctly taken into account by the equivalent hydraulic diameter. Useful information is also provided concerning the effect of a cold wall in an annulus.

  15. Computational analysis for prediction of pressure of PWR presurizer undertransient conditions

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    A computer model has been developed for prediction of the pressure in thepressurizer undertransient conditions. In the model three separate thermodynamic regions which arenot required to be inthermal equilibrium have been considered. The mathematical model derived from the general conservation equations includesall of theimportant thermal-hydraulics phenomena occurring in the pressurizer,i.e., stratificationof the hot water andincoming cold water, bulk flashing and condensation, wall condensation, andinterfacial heat and masstransfer, etc. The bubble rising and rain-out models are developed to describe bulkflashing andcondensation, respectively. To obtain the wall condensation rate, a one-dimensionalheat conductionequation is solved by the pivoting method. The presented model will predict thepressure-time behaviorof a PWR pressurizer during a variety of transients. The results obtained from the proposed mathematical model are in good agreementwithavailable data on the CHASHMA nuclear power plant's pressurizer performance.

  16. PRETTA:A COMPUTER PROGRAM FOR PWR PRESSURIZER’S TRANSIENT THERMODYNAMICS

    Institute of Scientific and Technical Information of China (English)

    阿谢德; 徐济鋆

    2001-01-01

    A computer program PRETTA “Pressurizer Transient Thermodynamics Analysis” was developed for the prediction of pressurizer under transient conditions. It is based on the solution of the conservation laws of heat and mass applied to the three separate and non equilibrium thermodynamic regions. In the program all of the important thermal-hydraulics phenomena occurring in the pressurizer: stratification of the hot water and incoming cold water, bulk flashing and condensation, wall condensation, and interfacial heat and mass transfer have been considered. The bubble rising and rain-out models are developed to describe bulk flashing and condensation, respectively. To obtain the wall condensation rate, a one-dimensional heat conduction equation is solved by the pivoting method. The presented computer program will predict the pressure-time behavior of a PWR pressurizer during a variety of transients. The results obtained from the proposed mathematical model are in good agreement with available data on the CHASHMA nuclear power plant's pressurizer performance.

  17. Demonstration of Uncertainty Quantification and Sensitivity Analysis for PWR Fuel Performance with BISON

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Hongbin; Ladd, Jacob; Zhao, Haihua; Zou, Ling; Burns, Douglas

    2015-11-01

    BISON is an advanced fuels performance code being developed at Idaho National Laboratory and is the code of choice for fuels performance by the U.S. Department of Energy (DOE)’s Consortium for Advanced Simulation of Light Water Reactors (CASL) Program. An approach to uncertainty quantification and sensitivity analysis with BISON was developed and a new toolkit was created. A PWR fuel rod model was developed and simulated by BISON, and uncertainty quantification and sensitivity analysis were performed with eighteen uncertain input parameters. The maximum fuel temperature and gap conductance were selected as the figures of merit (FOM). Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis.

  18. Pressure loss tests for DR-BEP of fullsize 17 x 17 PWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Moon Ki; Chun, Se Young; Chang, Seok Kyu; Won, Soon Youn; Cho, Young Rho; Kim, Bok Deuk; Min, Kyoung Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-01-01

    This report describes the conditions, procedure and results in the pressure loss tests carried out for a double grid type debris resistance bottom end piece (DR-BEP) designed by KAERI. In this test, the pressure loss coefficients of the full size 17 x 17 PWR simulated fuel assembly with DR-BET and with standard-BEP were measured respectively, and the pressure loss coefficients of DR-BEP were compared with the coefficients of STD-BET. The test conditions fall within the ranges of loop pressure from 5.2 to 45 bar, loop temperature from 27 to 221 deg C and Reynolds number in fuel bundle from 2.17 x 10{sup 4} to 3.85 x 10{sup 5}. (Author) 5 refs., 18 figs., 5 tabs.

  19. Common cause evaluations in applied risk analysis of nuclear power plants. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Taniguchi, T.; Ligon, D.; Stamatelatos, M.

    1983-04-01

    Qualitative and quantitative approaches were developed for the evaluation of common cause failures (CCFs) in nuclear power plants and were applied to the analysis of the auxiliary feedwater systems of several pressurized water reactors (PWRs). Key CCF variables were identified through a survey of experts in the field and a review of failure experience in operating PWRs. These variables were classified into categories of high, medium, and low defense against a CCF. Based on the results, a checklist was developed for analyzing CCFs of systems. Several known techniques for quantifying CCFs were also reviewed. The information provided valuable insights in the development of a new model for estimating CCF probabilities, which is an extension of and improvement over the Beta Factor method. As applied to the analysis of the PWR auxiliary feedwater systems, the method yielded much more realistic values than the original Beta Factor method for a one-out-of-three system.

  20. Revised uranium--plutonium cycle PWR and BWR models for the ORIGEN computer code

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A. G.; Bjerke, M. A.; Morrison, G. W.; Petrie, L. M.

    1978-09-01

    Reactor physics calculations and literature searches have been conducted, leading to the creation of revised enriched-uranium and enriched-uranium/mixed-oxide-fueled PWR and BWR reactor models for the ORIGEN computer code. These ORIGEN reactor models are based on cross sections that have been taken directly from the reactor physics codes and eliminate the need to make adjustments in uncorrected cross sections in order to obtain correct depletion results. Revised values of the ORIGEN flux parameters THERM, RES, and FAST were calculated along with new parameters related to the activation of fuel-assembly structural materials not located in the active fuel zone. Recommended fuel and structural material masses and compositions are presented. A summary of the new ORIGEN reactor models is given.

  1. Transient fuel behavior of preirradiated PWR fuels under reactivity initiated accident conditions

    Science.gov (United States)

    Fujishiro, Toshio; Yanagisawa, Kazuaki; Ishijima, Kiyomi; Shiba, Koreyuki

    1992-06-01

    Since 1975, extensive studies on transient fuel behavior under reactivity initiated accident (RIA) conditions have been continued in the Nuclear Safety Research Reactor (NSRR) of Japan Atomic Energy Research Institute. A new experimental program with preirradiated LWR fuel rods as test samples has recently been started. In this program, transient behavior and failure initiation have been studied with 14 × 14 type PWR fuel rods preirradiated to a burnup of 20 to 42 MWd/kgU. The test fuel rods contained in a capsule filled with the coolant water were subjected to a pulse irradiation in the NSRR to simulate a prompt power surge in an RIA. The effects of preirradiation on the transient fission gas release, pellet-cladding mechanical interaction and fuel failure were clearly observed through the transient in-core measurements and postirradiation examination.

  2. Quantitative uncertainty and sensitivity analysis of a PWR control rod ejection accident

    Energy Technology Data Exchange (ETDEWEB)

    Pasichnyk, I.; Perin, Y.; Velkov, K. [Gesellschaft flier Anlagen- und Reaktorsicherheit - GRS mbH, Boltzmannstasse 14, 85748 Garching bei Muenchen (Germany)

    2013-07-01

    The paper describes the results of the quantitative Uncertainty and Sensitivity (U/S) Analysis of a Rod Ejection Accident (REA) which is simulated by the coupled system code ATHLET-QUABOX/CUBBOX applying the GRS tool for U/S analysis SUSA/XSUSA. For the present study, a UOX/MOX mixed core loading based on a generic PWR is modeled. A control rod ejection is calculated for two reactor states: Hot Zero Power (HZP) and 30% of nominal power. The worst cases for the rod ejection are determined by steady-state neutronic simulations taking into account the maximum reactivity insertion in the system and the power peaking factor. For the U/S analysis 378 uncertain parameters are identified and quantified (thermal-hydraulic initial and boundary conditions, input parameters and variations of the two-group cross sections). Results for uncertainty and sensitivity analysis are presented for safety important global and local parameters. (authors)

  3. Degraded core accidents for the Sizewell PWR A sensitivity analysis of the radiological consequences

    CERN Document Server

    Kelly, G N; Clarke, R H; Ferguson, L; Haywood, S M; Hemming, C R; Jones, J A

    1982-01-01

    The radiological impact of degraded core accidents postulated for the Sizewell PWR was assessed in an earlier study. In this report the sensitivity of the predicted consequences to variation in the values of a number of important parameters is investigated for one of the postulated accidental releases. The parameters subjected to sensitivity analyses are the dose-mortality relationship for bone marrow irradiation, the energy content of the release, the warning time before the release to the environment, and the dry deposition velocity for airborne material. These parameters were identified as among the more important in determining the uncertainty in the results obtained in the initial study. With a few exceptions the predicted consequences were found to be not very sensitive to the parameter values investigated, the range of variation in the consequences for the limiting values of each parameter rarely exceeded a factor of a few and in many cases was considerably less. The conclusions reached are, however, p...

  4. The radiological impact on the Greater London population of postulated accidental releases from the Sizewell PWR

    CERN Document Server

    Kelly, G N; Charles, D; Hemming, C R

    1983-01-01

    This report contains an assessment of the radiological impact on the Greater London population of postulated accidental releases from the Sizewell PWR. Three of the degraded core accident releases postulated by the CEGB are analysed. The consequences, conditional upon each release, are evaluated in terms of the health impact on the exposed population and the impact of countermeasures taken to limit the exposure. Consideration is given to the risk to the Greater London population as a whole and to individuals within it. The consequences are evaluated using the NRPB code MARC (Methodology for Assessing Radiological Consequences). The results presented in this report are all conditional upon the occurrence of each release. In assessing the significance of the results, due account must be taken of the frequency with which such releases may be predicted to occur.

  5. Numerical modeling of in-vessel melt water interaction in large scale PWR`s

    Energy Technology Data Exchange (ETDEWEB)

    Kolev, N.I. [Siemens AG, KWU NA-M, Erlangen (Germany)

    1998-01-01

    This paper presents a comparison between IVA4 simulations and FARO L14, L20 experiments. Both experiments were performed with the same geometry but under different initial pressures, 51 and 20 bar respectively. A pretest prediction for test L21 which is intended to be performed under an initial pressure of 5 bar is also presented. The strong effect of the volume expansion of the evaporating water at low pressure is demonstrated. An in-vessel simulation for a 1500 MW el. PWR is presented. The insight gained from this study is: that at no time are conditions for the feared large scale melt-water intermixing at low pressure in force, with this due to the limiting effect of the expansion process which accelerates the melt and the water into all available flow paths. (author)

  6. An Extension of the Validation of SCALE (SAS2H) Isotopic Predictions for PWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1993-01-01

    Isotopic characterization of spent fuel via depletion and decay calculations is necessary for determination of source terms for subsequent system analyses involving heat transfer, radiation shielding, isotopic migration, etc. Unlike fresh fuel assumptions typically employed in the criticality safety analysis of spent fuel configurations, burnup credit applications also rely on depletion and decay calculations to predict the isotopic composition of spent fuel. These isotopics are used in subsequent criticality calculations to assess the reduced worth of spent fuel. To validate the codes and data used in depletion approaches, experimental measurements are compared with numerical predictions for relevant spent fuel samples. Such comparisons have been performed in earlier work at the Oak Ridge National Laboratory (ORNL). This report describes additional independent measurements and corresponding calculations, which supplement the results of the earlier work. The current work includes measured isotopic data from 19 spent fuel samples obtained from the Italian Trino Vercelles pressurized-water reactor (PWR) and the U.S. Turkey Point Unit 3 PWR. In addition, an approach to determine biases and uncertainties between calculated and measured isotopic concentrations is discussed, together with a method to statistically combine these terms to obtain a conservative estimate of spent fuel isotopic concentrations. Results are presented based on the combination of measured-to-calculated ratios for earlier work and the current analyses. The results described herein represent an extension to a new reactor design not included in the earlier work, and spent fuel samples with enrichment as high as 3.9 wt % {sup 235}U. Results for the current work are found to be, for the most part, consistent with the findings of the earlier work. This consistency was observed for results obtained from each of two different cross-section libraries and suggests that the estimated biases determined for

  7. Developing and analyzing long-term fuel management strategies for an advanced Small Modular PWR

    Energy Technology Data Exchange (ETDEWEB)

    Hedayat, Afshin, E-mail: ahedayat@aeoi.org.ir

    2017-03-15

    Highlights: • Comprehensive introduction and supplementary concepts as a review paper. • Developing an integrated long-term fuel management strategy for a SMR. • High reliable 3-D core modeling over fuel pins against the traditional LRM. • Verifying the expert rules of large PWRs for an advanced small PWR. • Investigating large numbers of safety parameters coherently. - Abstract: In this paper, long-term fuel management (FM) strategies are introduced and analyzed for a new advanced Pressurized Light Water Reactor (PWR) type of Small Modular Reactors (SMRs). The FM strategies are developed to be safe and practical for implementation as much as possible. Safety performances, economy of fuel, and Quality Assurance (QA) of periodic equilibrium conditions are chosen as the main goals. Flattening power density distribution over fuel pins is the major method to ensure safety performance; also maximum energy output or permissible discharging burn up indicates economy of fuel fabrication costs. Burn up effects from BOC to EOC have been traced, studied, and highly visualized in both of transport lattice cell calculations and diffusion core calculations. Long-term characteristics are searched to gain periodical equilibrium characteristics. They are fissile changes, neutron spectrum, refueling pattern, fuel cycle length, core excess reactivity, average, and maximum burn up of discharged fuels, radial Power Peaking Factors (PPF), total PPF, radial and axial power distributions, batch effects, and enrichment effects for fine regulations. Traditional linear reactivity model have been successfully simulated and adapted via fine core and burn up calculations. Effects of high burnable neutron poison and soluble boron are analyzed. Different numbers of batches via different refueling patterns have been studied and visualized. Expert rules for large type PWRs have been influenced and well tested throughout accurate equilibrium core calculations.

  8. Evaluation of Physical Characteristics of PWR Cores with Accident Tolerant Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hee; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of); In, Wang Kee [KAERI, Daejeon (Korea, Republic of)

    2015-10-15

    The accident tolerant fuels (ATF) considered in this work includes metallic microcell UO{sub 2} pellets and outer Cr-based alloy coating on cladding, which is being developed in KAERI (Korea Atomic Energy Research Institute). Chromium metals have been used in many fields because of its hardness and corrosion-resistance. The use of the chromium metal in nuclear fuel rod can enhance the conductivity of pellets and corrosion-resistance of cladding. The objective of this work is to study the neutronic performances and characteristics of the commercial PWR core loaded the ATF-bearing assemblies. In this work, we studied the PWR cores which are loaded with ATF assemblies to improve the safety of reactor core. The ATF rod consists of the metallic microcell UO2 pellet which includes chromium of 3.34 wt% and the outer 0.05mm thick coating of Cr-based alloy with atomic number ratio of 85:15. We performed the cycle-by-cycle reload core analysis from the cycle 8 at which the ATF fuel assemblies start to be loaded into the core. The target nuclear power plant is the Hanbit-3 nuclear power plant. From the analysis, it was found that 1) the uranium enrichment is required to be increased up to 5.20/4.70 wt% in order to satisfy a required cycle length of 480 EFPDs, 2) the cycle length for the core using ATF fuel assemblies with the same uranium enrichments as those in the reference UO{sub 2} fueled core is decreased from 480 EFPDs to 430 EFPDs.

  9. Singular deposit formation in PWR due to electrokinetic phenomena - application to SG clogging

    Energy Technology Data Exchange (ETDEWEB)

    Guillodo, M.; Muller, T.; Barale, M.; Foucault, M. [AREVA NP SAS, Technical Centre (France); Clinard, M.-H.; Brun, C.; Chahma, F. [AREVA NP SAS, Chemistry and Radiochemistry Group (France); Corredera, G.; De Bouvier, O. [Electricite de France, Centre d' Expertise de I' inspection dans les domaines de la Realisation et de l' Exploitation (France)

    2009-07-01

    The deposits which cause clogging of the 'foils' of the tube support plates (TSP) in Steam Generators (SG) of PWR present two characteristics which put forward that the mechanism at the origin of their formation is different from the mechanism that drives the formation of homogeneous deposits leading to the fouling of the free spans of SG tubes. Clogging occurs near the leading edge of the TSP and the deposits appear as diaphragms localized between both TSP and SG tubing materials, while the major part of the tube/TSP interstice presents little or no significant clogging. This type of deposit seems rather comparable to the ones which were reproduced in Lab tests to explain the flow rate instabilities observed on a French unit during hot shutdown in the 90's. The deposits which cause TSP clogging are owed to a discontinuity of the streaming currents in the vicinity of a surface singularity (orifices, scratches ...) which, in very low conductivity environment, produce local potential variations and/or current loop in the metallic pipe material due to electrokinetic effects. Deposits can be built by two mechanisms which may or not coexist: (i) accumulation of particles stabilized by an electrostatic attraction due to the local variation of electrokinetic potential, and (ii) crystalline growth of magnetite produced by the oxidation of ferrous ions on the anodic branch of a current loop. Lab investigations carried out by AREVA NP Technical Centre since the end of the 90's showed that this type of deposit occurs when the redox potential is higher than a critical value, and can be gradually dissolved when the potential becomes lower than this value which depends on the 'Material - Chemistry' couple. Special emphasis will be given in this paper to the TSP clogging of SG in PWR secondary coolant dealing particularly with the potential strong effect of electrokinetic phenomena in low conductive environment and in high temperature conditions

  10. Computational fluid dynamics (CFD) round robin benchmark for a pressurized water reactor (PWR) rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Shin K., E-mail: paengki1@tamu.edu; Hassan, Yassin A.

    2016-05-15

    Highlights: • The capabilities of steady RANS models were directly assessed for full axial scale experiment. • The importance of mesh and conjugate heat transfer was reaffirmed. • The rod inner-surface temperature was directly compared. • The steady RANS calculations showed a limitation in the prediction of circumferential distribution of the rod surface temperature. - Abstract: This study examined the capabilities and limitations of steady Reynolds-Averaged Navier–Stokes (RANS) approach for pressurized water reactor (PWR) rod bundle problems, based on the round robin benchmark of computational fluid dynamics (CFD) codes against the NESTOR experiment for a 5 × 5 rod bundle with typical split-type mixing vane grids (MVGs). The round robin exercise against the high-fidelity, broad-range (covering multi-spans and entire lateral domain) NESTOR experimental data for both the flow field and the rod temperatures enabled us to obtain important insights into CFD prediction and validation for the split-type MVG PWR rod bundle problem. It was found that the steady RANS turbulence models with wall function could reasonably predict two key variables for a rod bundle problem – grid span pressure loss and the rod surface temperature – once mesh (type, resolution, and configuration) was suitable and conjugate heat transfer was properly considered. However, they over-predicted the magnitude of the circumferential variation of the rod surface temperature and could not capture its peak azimuthal locations for a central rod in the wake of the MVG. These discrepancies in the rod surface temperature were probably because the steady RANS approach could not capture unsteady, large-scale cross-flow fluctuations and qualitative cross-flow pattern change due to the laterally confined test section. Based on this benchmarking study, lessons and recommendations about experimental methods as well as CFD methods were also provided for the future research.

  11. Analysis of nuclear characteristics and fuel economics for PWR core with homogeneous thorium fuels

    Energy Technology Data Exchange (ETDEWEB)

    Joo, H. K.; Noh, J. M.; Yoo, J. W.; Song, J. S.; Kim, J. C.; Noh, T. W

    2000-12-01

    The nuclear core characteristics and economics of an once-through homogenized thorium cycle for PWR were analyzed. The lattice code, HELIOS has been qualified against BNL and B and W critical experiments and the IAEA numerical benchmark problem in advance of the core analysis. The infinite multiplication factor and the evolution of main isotopes with fuel burnup were investigated for the assessment of depletion charateristics of thorium fuel. The reactivity of thorium fuel at the beginning of irradiation is smaller than that of uranium fuel having the same inventory of {sup 235}U, but it decrease with burnup more slowly than in UO{sub 2} fuel. The gadolinia worth in thorium fuel assembly is also slightly smaller than in UO{sub 2} fuel. The inventory of {sup 233}U which is converted from {sup 232}Th is proportional to the initial mass of {sup 232}Th and is about 13kg per one tones of initial heavy metal mass. The followings are observed for thorium fuel cycle compared with UO{sub 2} cycle ; shorter cycle length, more positive MTC at EOC, more negative FTC, similar boron worth and control rod. Fuel economics of thorium cycle was analyzed by investigating the natural uranium requirements, the separative work requirements, and the cost for burnable poison rods. Even though less number of burnable poison rods are required in thorium fuel cycle, the costs for the natural uranium requirements and the separative work requirements are increased in thorium fuel cycle. So within the scope of this study, once through cycle concept, homogenized fuel concept, the same fuel management scheme as uranium cycle, the thorium fuel cycle for PWR does not have any economic incentives in preference to uranium.

  12. Effect of aging on the PWR Chemical and Volume Control System

    Energy Technology Data Exchange (ETDEWEB)

    Grove, E.J.; Travis, R.J.; Aggarwal, S.K. [Brookhaven National Lab., Upton, NY (United States)

    1995-06-01

    The PWR Chemical and Volume Control System (CVCS) is designed to provide both safety and non-safety related functions. During normal plant operation it is used to control reactor coolant chemistry, and letdown and charging flow. In many plants, the charging pumps also provide high pressure injection, emergency boration, and RCP seal injection in emergency situations. This study examines the design, materials, maintenance, operation and actual degradation experiences of the system and main sub-components to assess the potential for age degradation. A detailed review of the Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Report (LER) databases for the 1988--1991 time period, together with a review of industry and NRC experience and research, indicate that age-related degradations and failures have occurred. These failures had significant effects on plant operation, including reactivity excursions, and pressurizer level transients. The majority of these component failures resulted in leakage of reactor coolant outside the containment. A representative plant of each PWR design (W, CE, and B and W) was visited to obtain specific information on system inspection, surveillance, monitoring, and inspection practices. The results of these visits indicate that adequate system maintenance and inspection is being performed. In some instances, the frequencies of inspection were increase in response to repeated failure events. A parametric study was performed to assess the effect of system aging on Core Damage Frequency (CDF). This study showed that as motor-operated valve (MOV) operating failures increased, the contribution of the High Pressure Injection to CDF also increased.

  13. Second sector cool down

    CERN Multimedia

    2007-01-01

    At the beginning of July, cool-down is starting in the second LHC sector, sector 4-5. The cool down of sector 4-5 may occasionally generate mist at Point 4, like that produced last January (photo) during the cool-down of sector 7-8.Things are getting colder in the LHC. Sector 7-8 has been kept at 1.9 K for three weeks with excellent stability (see Bulletin No. 16-17 of 16 April 2007). The electrical tests in this sector have got opt to a successful start. At the beginning of July the cryogenic teams started to cool a second sector, sector 4-5. At Point 4 in Echenevex, where one of the LHC’s cryogenic plants is located, preparations for the first phase of the cool-down are underway. During this phase, the sector will first be cooled to 80 K (-193°C), the temperature of liquid nitrogen. As for the first sector, 1200 tonnes of liquid nitrogen will be used for the cool-down. In fact, the nitrogen circulates only at the surface in the ...

  14. Measuring the coolness of interactive products: the COOL questionnaire

    DEFF Research Database (Denmark)

    Bruun, Anders; Raptis, Dimitrios; Kjeldskov, Jesper;

    2016-01-01

    is the COOL questionnaire. We based the creation of the questionnaire on literature suggesting that perceived coolness is decomposed to outer cool (the style of a product) and inner cool (the personality characteristics assigned to it). In this paper, we focused on inner cool, and we identified 11 inner cool......, rebelliousness and usability. These factors and their underlying 16 question items comprise the COOL questionnaire. The whole process of creating the questionnaire is presented in detail in this paper and we conclude by discussing our work against related work on coolness and HCI....

  15. Progress and prospects of nuclear fuel development in Japan, (2). Progress and future plan of research and development on PWR fuel in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kondo, Yoshiaki; Abeta, Sadaaki; Aisu, Hideo; Teranishi, Tomoyuki

    1982-06-01

    13 years have elapsed since the first PWR plant started the operation in Japan, and at present, 11 PWR plants are in operation. During this period, much results of use and experience have been accumulated for the PWR fuel. The improvement and development of the fuel have been performed to meet the supply of the fuel sufficiently adaptable to the severe environment in Japan. In this paper, the evaluation of soundness and the improvement of reliability of PWR fuel made so far are reported, and the response of fuel side to long cycle operation and load following-up operation, which will be required in near future, is explained. The inspection of fuel has been performed at reactor sites for the purpose of sufficiently observing the irradiation behavior of fuel and detecting the points out of order. Effort has been exerted to perform various inspections thoroughly on total number of fuel and reflect the results to the improved design. Fuel leak scarcely occurred from the beginning, accordingly, improvement has been made to reduce the bending of fuel rods. The change of PWR fuel design, the evaluation of soundness and the improvement of reliability of PWR fuel, and the improvement for the future are reported.

  16. Cool WISPs for stellar cooling excesses

    Energy Technology Data Exchange (ETDEWEB)

    Giannotti, Maurizio [Barry Univ., Miami Shores, FL (United States). Physical Sciences; Irastorza, Igor [Zaragoza Univ. (Spain). Dept. de Fisica Teorica; Redondo, Javier [Zaragoza Univ. (Spain). Dept. de Fisica Teorica; Max-Planck-Institut fuer Physik, Muenchen (Germany); Ringwald, Andreas [DESY Hamburg (Germany). Theory Group

    2015-12-15

    Several stellar systems (white dwarfs, red giants, horizontal branch stars and possibly the neutron star in the supernova remnant Cassiopeia A) show a preference for a mild non-standard cooling mechanism when compared with theoretical models. This exotic cooling could be provided by Weakly Interacting Slim Particles (WISPs), produced in the hot cores and abandoning the star unimpeded, contributing directly to the energy loss. Taken individually, these excesses do not show a strong statistical weight. However, if one mechanism could consistently explain several of them, the hint could be significant. We analyze the hints in terms of neutrino anomalous magnetic moments, minicharged particles, hidden photons and axion-like particles (ALPs). Among them, the ALP represents the best solution. Interestingly, the hinted ALP parameter space is accessible to the next generation proposed ALP searches, such as ALPS II and IAXO.

  17. Cool WISPs for stellar cooling excesses

    CERN Document Server

    Giannotti, Maurizio; Redondo, Javier; Ringwald, Andreas

    2015-01-01

    Several stellar systems (white dwarfs, red giants, horizontal branch stars and possibly the neutron star in the supernova remnant Cassiopeia A) show a preference for a mild non-standard cooling mechanism when compared with theoretical models. This exotic cooling could be provided by Weakly Interacting Slim Particles (WISPs), produced in the hot cores and abandoning the star unimpeded, contributing directly to the energy loss. Taken individually, these excesses do not show a strong statistical weight. However, if one mechanism could consistently explain several of them, the hint could be significant. We analyze the hints in terms of neutrino anomalous magnetic moments, minicharged particles, hidden photons and axion-like particles (ALPs). Among them, the ALP represents the best solution. Interestingly, the hinted ALP parameter space is accessible to the next generation proposed ALP searches, such as ALPS II and IAXO.

  18. Water-cooled electronics

    CERN Document Server

    Dumont, G; Righini, B

    2000-01-01

    LHC experiments demand on cooling of electronic instrumentation will be extremely high. A large number of racks will be located in underground caverns and counting rooms, where cooling by conventional climatisation would be prohibitively expensive. A series of tests on the direct water cooling of VMEbus units and of their standard power supplies is reported. A maximum dissipation of 60 W for each module and more than 1000 W delivered by the power supply to the crate have been reached. These values comply with the VMEbus specifications. (3 refs).

  19. Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP; Calculos de criticidad de un contenedor de combustible generico para ensambles combustibles PWR, mediante el codigo MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Vargas E, S.; Esquivel E, J.; Ramirez S, J. R., E-mail: samuel.vargas@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

  20. Effects of generation and optimization of libraries of effective sections in the analysis of transient in PWR reactors; Efectos de generacion y optimizacion de librerias de secciones eficaces en el analisis de transitorios en reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez-Cervera, S.; Garcia Herranz, N.; Cuervo, D.; Ahnert, C.

    2014-07-01

    In this paper evaluates the impact that has a certain mesh on a transient in a PWR reactor in the expulsion of a control bar. Have been used for this purpose the coupled codes neutronic and Thermo-hydraulic COBAYA3/COBRA-TF. This objective has been chosen the OECD/NEA PWR MOX/UO{sub 2} rod ejection transient benchmark provides isotopic compositions and defined geometric configurations that allow the use of codes lattice to generate own bookstores. The code used for this transport has been the code APOLLO2.8. The results show large discrepancies when using the benchmark library or libraries own by comparing them to the other participants solutions. The source of these discrepancies is the nodal effective sections provided in the benchmark. (Author)

  1. Utilization of waste heat from Vienna waste incinerators for the operation of a district cooling grid. Effects on the primary energy efficiency of district heating and district cooling in Vienna; Nutzung der Abwaerme aus den Wiener Abfallverbrennungsanlagen fuer den Betrieb eines Fernkaeltenetzes. Auswirkungen auf die Primaerenergieeffizienz der Fernwaerme und Fernkaelte in Wien

    Energy Technology Data Exchange (ETDEWEB)

    Schindelar, F.; Wallisch, A. [Fernwaerme Wien GmbH, Vienna (Austria)

    2007-07-01

    The need of coldness increases and has to be covered efficiently as well as ecologically. At optimal constellation and mode of operation, the establishment of refrigeration plants from absorption refrigerators and compression refrigerators seems to be economically more competitive than decentralized plants. The optimal constellation is present, if: (a) ecologically and economically favourable waste heat are available; (b) Electricity from the domestic production with waste energy is present; (c) Resources-conserving recirculation cooling possibilities exist; (d) cooling water tanks and/or hot water tanks are available for top coverage; (e) a high grid density exists; (f) in-building station corresponds to the technical conditions. If these fundamental conditions are present, then the district coldness offers a good chance for waste incineration plants to use a safe heat consumer also in summer and to utilize optimally the existing energy.

  2. Development of a model of a NSSS of the PWR reactor with thermo-hydraulic code GOTHIC; Desarrollo de un modelo del NSSS de un reactor PWR con el codigo termo-hidraulico GOTHIC

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Garcia-Torano, I.; Jimenez, G.

    2013-07-01

    The Thermo-hydraulic code GOTHIC is often used in the nuclear industry for licensing transient analysis inside containment of generation II (PWR, BWR) plants as Gen III and III + (AP1000, ESBWR, APWR). After entering the mass and energy released to the containment, previously calculated by other codes (basis, TRACE), GOTHIC allows to calculate in detail the evolution of basic parameters in the containment.

  3. Cooling Devices in Laser therapy.

    Science.gov (United States)

    Das, Anupam; Sarda, Aarti; De, Abhishek

    2016-01-01

    Cooling devices and methods are now integrated into most laser systems, with a view to protecting the epidermis, reducing pain and erythema and improving the efficacy of laser. On the basis of method employed, it can be divided into contact cooling and non-contact cooling. With respect to timing of irradiation of laser, the nomenclatures include pre-cooling, parallel cooling and post-cooling. The choice of the cooling device is dictated by the laser device, the physician's personal choice with respect to user-friendliness, comfort of the patient, the price and maintenance costs of the device. We hereby briefly review the various techniques of cooling, employed in laser practice.

  4. LHC cooling gains ground

    CERN Multimedia

    Huillet-Miraton Catherine

    The nominal cryogenic conditions of 1.9 K have been achieved in sectors 5-6 and 7-8. This means that a quarter of the machine has reached the nominal conditions for LHC operation, having attained a temperature of below 2 K (-271°C), which is colder than interstellar space! Elsewhere, the cryogenic system in Sector 8-1 has been filled with liquid helium and cooled to 2K and will soon be available for magnet testing. Sectors 6-7 and 2-3 are being cooled down and cool-down operations have started in Sector 3-4. Finally, preparations are in hand for the cool-down of Sector 1-2 in May and of Sector 4-5, which is currently being consolidated. The LHC should be completely cold for the summer. For more information: http://lhc.web.cern.ch/lhc/Cooldown_status.htm.

  5. Why Exercise Is Cool

    Science.gov (United States)

    ... to Know About Puberty Train Your Temper Why Exercise Is Cool KidsHealth > For Kids > Why Exercise Is ... day and your body will thank you later! Exercise Makes Your Heart Happy You may know that ...

  6. Waveguide cooling system

    Science.gov (United States)

    Chen, B. C. J.; Hartop, R. W.

    1981-04-01

    An improved system is described for cooling high power waveguides by the use of cooling ducts extending along the waveguide, which minimizes hot spots at the flanges where waveguide sections are connected together. The cooling duct extends along substantially the full length of the waveguide section, and each flange at the end of the section has a through hole with an inner end connected to the duct and an opposite end that can be aligned with a flange hole in another waveguide section. Earth flange is formed with a drainage groove in its face, between the through hole and the waveguide conduit to prevent leakage of cooling fluid into the waveguide. The ducts have narrowed sections immediately adjacent to the flanges to provide room for the installation of fasteners closely around the waveguide channel.

  7. Warm and Cool Dinosaurs.

    Science.gov (United States)

    Mannlein, Sally

    2001-01-01

    Presents an art activity in which first grade students draw dinosaurs in order to learn about the concept of warm and cool colors. Explains how the activity also helped the students learn about the concept of distance when drawing. (CMK)

  8. Cooling of wood briquettes

    Directory of Open Access Journals (Sweden)

    Adžić Miroljub M.

    2013-01-01

    Full Text Available This paper is concerned with the experimental research of surface temperature of wood briquettes during cooling phase along the cooling line. The cooling phase is an important part of the briquette production technology. It should be performed with care, otherwise the quality of briquettes could deteriorate and possible changes of combustion characteristics of briquettes could happen. The briquette surface temperature was measured with an IR camera and a surface temperature probe at 42 sections. It was found that the temperature of briquette surface dropped from 68 to 34°C after 7 minutes spent at the cooling line. The temperature at the center of briquette, during the 6 hour storage, decreased to 38°C.

  9. Stacking with stochastic cooling

    Energy Technology Data Exchange (ETDEWEB)

    Caspers, Fritz E-mail: Fritz.Caspers@cern.ch; Moehl, Dieter

    2004-10-11

    Accumulation of large stacks of antiprotons or ions with the aid of stochastic cooling is more delicate than cooling a constant intensity beam. Basically the difficulty stems from the fact that the optimized gain and the cooling rate are inversely proportional to the number of particles 'seen' by the cooling system. Therefore, to maintain fast stacking, the newly injected batch has to be strongly 'protected' from the Schottky noise of the stack. Vice versa the stack has to be efficiently 'shielded' against the high gain cooling system for the injected beam. In the antiproton accumulators with stacking ratios up to 10{sup 5} the problem is solved by radial separation of the injection and the stack orbits in a region of large dispersion. An array of several tapered cooling systems with a matched gain profile provides a continuous particle flux towards the high-density stack core. Shielding of the different systems from each other is obtained both through the spatial separation and via the revolution frequencies (filters). In the 'old AA', where the antiproton collection and stacking was done in one single ring, the injected beam was further shielded during cooling by means of a movable shutter. The complexity of these systems is very high. For more modest stacking ratios, one might use azimuthal rather than radial separation of stack and injected beam. Schematically half of the circumference would be used to accept and cool new beam and the remainder to house the stack. Fast gating is then required between the high gain cooling of the injected beam and the low gain stack cooling. RF-gymnastics are used to merge the pre-cooled batch with the stack, to re-create free space for the next injection, and to capture the new batch. This scheme is less demanding for the storage ring lattice, but at the expense of some reduction in stacking rate. The talk reviews the 'radial' separation schemes and also gives some

  10. Qualitative analysis of the maintenance politics of the systems of a typical PWR by artificial neural networks; Analise qualitativa da politica de manutencoes dos sistemas de um PWR tipico por redes neurais artificiais

    Energy Technology Data Exchange (ETDEWEB)

    Lourenco, Victor Hugo Moreno

    2010-02-15

    Proceedings and techniques in order to maximize the reliability and the availability of industrial plants have been used along the last decades by specialists and professionals of maintenance. However, the modem industrial systems' sizing, and the increasing complexity and interdependence among its components have become this activity's planning a more and more difficult task. Considering this scenario, the objective of the present work is to provide a computational tool which is able to help about the taking decision's task, and about planning policies of maintenance practiced in thermonuclear plants. The tool developed is based on the artificial neural networks (ANN) for the recognition of standards and establishment of correlations among events occurred in the components of pressurized water reactor (PWR) typical systems. The ANN work as miners of database of failure events, and are able to identify connections and to establish imperceptible inferences even for the most experienced specialists in maintenance of nuclear systems. The results were attained from realistic data and are confronted against the maintenance's classic policies which are practiced nowadays on PWR thermonuclear plants. These results show the solidity of the technique in valuing and predicting failures in a real power plant, and is able to be used as a tool for supporting decisions about planning maintenance policies on a typical PWR. (author)

  11. Cooling of electronic equipment

    DEFF Research Database (Denmark)

    A. Kristensen, Anders Schmidt

    2003-01-01

    Cooling of electronic equipment is studied. The design size of electronic equipment decrease causing the thermal density to increase. This affect the cooling which can cause for example failures of critical components due to overheating or thermal induced stresses. Initially a pin fin heat sink...... is considered as extruded profiles are inadequate for compact designs. An optimal pin fin shape and configuration is sought also taking manufacturing costs into consideration. Standard methods for geometrical modeling and thermal analysis are applied....

  12. Anomalous law of cooling

    OpenAIRE

    Lapas, Luciano C.; Ferreira, Rogelma M. S.; Oliveira, Fernando A.; Rubí, J. Miguel

    2014-01-01

    We analyze the temperature relaxation phenomena of systems in contact with a thermal reservoir that undergo a non-Markovian diffusion process. From a generalized Langevin equation, we show that the temperature is governed by a law of cooling of the Newton's law type in which the relaxation time depends on the velocity autocorrelation and is then characterized by the memory function. The analysis of the temperature decay reveals the existence of an anomalous cooling in which the temperature ma...

  13. Cooling tower waste reduction

    Energy Technology Data Exchange (ETDEWEB)

    Coleman, S.J.; Celeste, J.; Chine, R.; Scott, C.

    1998-05-01

    At Lawrence Livermore National Laboratory (LLNL), the two main cooling tower systems (central and northwest) were upgraded during the summer of 1997 to reduce the generation of hazardous waste. In 1996, these two tower systems generated approximately 135,400 lbs (61,400 kg) of hazardous sludge, which is more than 90 percent of the hazardous waste for the site annually. At both, wet decks (cascade reservoirs) were covered to block sunlight. Covering the cascade reservoirs reduced the amount of chemical conditioners (e.g. algaecide and biocide), required and in turn the amount of waste generated was reduced. Additionally, at the northwest cooling tower system, a sand filtration system was installed to allow cyclical filtering and backflushing, and new pumps, piping, and spray nozzles were installed to increase agitation. the appurtenance upgrade increased the efficiency of the cooling towers. The sand filtration system at the northwest cooling tower system enables operators to continuously maintain the cooling tower water quality without taking the towers out of service. Operational costs (including waste handling and disposal) and maintenance activities are compared for the cooling towers before and after upgrades. Additionally, the effectiveness of the sand filter system in conjunction with the wet deck covers (northwest cooling tower system), versus the cascade reservoir covers alone (south cooling tower south) is discussed. the overall expected return on investment is calculated to be in excess of 250 percent. this upgrade has been incorporated into the 1998 DOE complex-wide water conservation project being led by Sandia National Laboratory/Albuquerque.

  14. Cooling with Superfluid Helium

    CERN Document Server

    Lebrun, P

    2014-01-01

    The technical properties of helium II (‘superfluid’ helium) are presented in view of its applications to the cooling of superconducting devices, particularly in particle accelerators. Cooling schemes are discussed in terms of heat transfer performance and limitations. Large-capacity refrigeration techniques below 2 K are reviewed, with regard to thermodynamic cycles as well as process machinery. Examples drawn from existing or planned projects illustrate the presentation. Keywords: superfluid helium, cryogenics

  15. Laser cooling of solids

    Energy Technology Data Exchange (ETDEWEB)

    Epstein, Richard I [Los Alamos National Laboratory; Sheik-bahae, Mansoor [UNM

    2008-01-01

    We present an overview of solid-state optical refrigeration also known as laser cooling in solids by fluorescence upconversion. The idea of cooling a solid-state optical material by simply shining a laser beam onto it may sound counter intuitive but is rapidly becoming a promising technology for future cryocooler. We chart the evolution of this science in rare-earth doped solids and semiconductors.

  16. Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-1 and TSE-2

    Energy Technology Data Exchange (ETDEWEB)

    Cheverton, R.D.

    1976-09-01

    The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and two thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. The PWR calculations indicated that under some circumstances crack propagation could be expected and that experiments should be conducted for cracks that would have the potential for propagation at least halfway through the wall.

  17. Application of RELAP5/MOD1 for calculation of safety and relief valve discharge piping hydrodynamic loads. Final report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    1982-12-01

    A series of operability tests of spring-loaded safety valves was performed at Combustion Engineering in Windsor, CT as part of the PWR Safety and Relief Valve Test Program conducted by EPRI on behalf of PWR Utilities in response to the recommendations of NUREG-0578 and the requirements of the NRC. Experimental data from five of the safety valve tests are compared with RELAP5/MOD1 calculations to evaluate the capability of the code to determine the fluid-induced transient loads on downstream piping. Comparisons between data and calculations are given for transients with discharge of steam, water, and water loop seal followed by steam. RELAP5/MOD1 provides useful engineering estimates of the fluid-induced piping loads for all cases.

  18. Calculation of sample problems related to two-phase flow blowdown transients in pressure relief piping of a PWR pressurizer

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Y.W.; Wiedermann, A.H.

    1984-02-01

    A method was published, based on the integral method of characteristics, by which the junction and boundary conditions needed in computation of a flow in a piping network can be accurately formulated. The method for the junction and boundary conditions formulation together with the two-step Lax-Wendroff scheme are used in a computer program; the program in turn, is used here in calculating sample problems related to the blowdown transient of a two-phase flow in the piping network downstream of a PWR pressurizer. Independent, nearly exact analytical solutions also are obtained for the sample problems. Comparison of the results obtained by the hybrid numerical technique with the analytical solutions showed generally good agreement. The good numerical accuracy shown by the results of our scheme suggest that the hybrid numerical technique is suitable for both benchmark and design calculations of PWR pressurizer blowdown transients.

  19. Comparative analysis between measured and calculated concentrations of major actinides using destructive assay data from Ohi-2 PWR

    Directory of Open Access Journals (Sweden)

    Oettingen Mikołaj

    2015-09-01

    Full Text Available In the paper, we assess the accuracy of the Monte Carlo continuous energy burnup code (MCB in predicting final concentrations of major actinides in the spent nuclear fuel from commercial PWR. The Ohi-2 PWR irradiation experiment was chosen for the numerical reconstruction due to the availability of the final concentrations for eleven major actinides including five uranium isotopes (U-232, U-234, U-235, U-236, U-238 and six plutonium isotopes (Pu-236, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242. The main results were presented as a calculated-to-experimental ratio (C/E for measured and calculated final actinide concentrations. The good agreement in the range of ±5% was obtained for 78% C/E factors (43 out of 55. The MCB modeling shows significant improvement compared with the results of previous studies conducted on the Ohi-2 experiment, which proves the reliability and accuracy of the developed methodology.

  20. Stacking with Stochastic Cooling

    CERN Document Server

    Caspers, Friedhelm

    2004-01-01

    Accumulation of large stacks of antiprotons or ions with the aid of stochastic cooling is more delicate than cooling a constant intensity beam. Basically the difficulty stems from the fact that the optimized gain and the cooling rate are inversely proportional to the number of particles seen by the cooling system. Therefore, to maintain fast stacking, the newly injected batch has to be strongly protected from the Schottky noise of the stack. Vice versa the stack has to be efficiently shielded against the high gain cooling system for the injected beam. In the antiproton accumulators with stacking ratios up to 105, the problem is solved by radial separation of the injection and the stack orbits in a region of large dispersion. An array of several tapered cooling systems with a matched gain profile provides a continuous particle flux towards the high-density stack core. Shielding of the different systems from each other is obtained both through the spatial separation and via the revolution frequencies (filters)....

  1. Alternative Room Cooling System

    Directory of Open Access Journals (Sweden)

    Md. Fazle Rabbi

    2015-06-01

    Full Text Available The rapidly growing population results in an increasing demand for much more residential and commercial buildings, which leads to vertical growth of the buildings and needs proper ventilation of those buildings. Natural air ventilation system is not sufficient for conventional building structures. Hence fans and air-conditioners are must to meet the requirement of proper ventilation as well as space conditioning. Globally building sector consumes largest energy in heating, cooling, ventilation and space conditioning. This load can be minimized by the application of solar chimney and modification in building structure for heating, cooling, ventilation and space conditioning. Passive solar cooling is a subject of interest to provide cooling by using the sun, a powerful energy source. This is done for ensuring human comfort in hot climates. ASHRAE (American Society of Heating, Refrigerating and Air Conditioning Engineers defines Comfort as ‘that state of mind which expresses satisfaction with the thermal environment.’ The present paper describes the development of a solar passive cooling system, which can provide thermal cooling throughout the summer season in hot and humid climates. The constructed passive system works on natural convection mode of air. Such system reduces the inside temperature of up to 5°C from the atmospheric temperature. Temperature can further be reduced by the judicious use of night ventilation.

  2. SAS2H Generated Isotopic Concentrations For B&W 15X15 PWR Assembly (SCPB:N/A)

    Energy Technology Data Exchange (ETDEWEB)

    J.W. Davis

    1996-08-29

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide pressurized water reactor (PWR) isotopic composition data as a function of time for use in criticality analyses. The objectives of this evaluation are to generate burnup and decay dependant isotopic inventories and to provide these inventories in a form which can easily be utilized in subsequent criticality calculations.

  3. Stakes and Solutions for current and up-coming Licensing Challenges in PWR and BWR Reload and Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curca-Tiving, F.; Opel, S.

    2014-07-01

    Regulatory requirements for reloads and safety analyses are evolving: New safety criteria, requests for enlarged qualification databases, statistical applications, uncertainty propagation... In order to address these challenges and access more predictable licensing processes, AREVA implements a consistent code and methodology suite for PWR and BWR core design and safety analysis, based on a first principles modeling with an extremely broad international verification and validation data base. (Author)

  4. Comparing Social Stories™ to Cool versus Not Cool

    Science.gov (United States)

    Leaf, Justin B.; Mitchell, Erin; Townley-Cochran, Donna; McEachin, John; Taubman, Mitchell; Leaf, Ronald

    2016-01-01

    In this study we compared the cool versus not cool procedure to Social Stories™ for teaching various social behaviors to one individual diagnosed with autism spectrum disorder. The researchers randomly assigned three social skills to the cool versus not cool procedure and three social skills to the Social Stories™ procedure. Naturalistic probes…

  5. Influence of oxide films on primary water stress corrosion cracking initiation of alloy 600

    Science.gov (United States)

    Panter, J.; Viguier, B.; Cloué, J.-M.; Foucault, M.; Combrade, P.; Andrieu, E.

    2006-01-01

    In the present study alloy 600 was tested in simulated pressurised water reactor (PWR) primary water, at 360 °C, under an hydrogen partial pressure of 30 kPa. These testing conditions correspond to the maximum sensitivity of alloy 600 to crack initiation. The resulting oxidised structures (corrosion scale and underlying metal) were characterised. A chromium rich oxide layer was revealed, the underlying metal being chromium depleted. In addition, analysis of the chemical composition of the metal close to the oxide scale had allowed to detect oxygen under the oxide scale and particularly in a triple grain boundary. Implication of such a finding on the crack initiation of alloy 600 is discussed. Significant diminution of the crack initiation time was observed for sample oxidised before stress corrosion tests. In view of these results, a mechanism for stress corrosion crack initiation of alloy 600 in PWR primary water was proposed.

  6. ANALISIS SENSITIVITAS TURBULENSI ALIRAN PADA KANAL BAHAN BAKAR PWR BERBASIS CFD

    Directory of Open Access Journals (Sweden)

    Endiah Puji Hastuti

    2015-04-01

    Full Text Available Turbulensi aliran pendingin pada proses perpindahan panas berfungsi untuk meningkatkan nilai koefisien perpindahan panas, tidak terkecuali aliran dalam kanal bahan bakar. Program CFD (CFD=computational fluid dynamics, FLUENT adalah program komputasi berbasis elemen hingga (finite element yang mampu memprediksi dan menganalisis fenomena dinamika aliran fluida secara teliti. Program perhitungan CFD dipilih dalam penelitian ini karena selain akurat juga dapat memberikan visualisasi dengan baik. Penelitian ini bertujuan untuk memahami karakteristika perpindahan panas, massa dan momentum dari dinding rod bahan bakar ke pendingin secara visual, pada medan temperatur, medan tekanan, dan medan energi kinetika pendingin, sebagai fungsi dinamika aliran di dalam kanal, pada kondisi tunak dan transien. Analisis dinamika aliran pada kanal bahan bakar PWR berbasis CFD dilakukan dengan menggunakan sampel data reaktor PWR dengan daya 1000 MWe dengan susunan bahan bakar 17x17. Untuk menguji sensitivitas persamaan aliran yang sesuai dengan model aliran turbulen pada kanal bahan bakar dilakukan pemodelan dengan menggunakan persamaan k-omega (Ƙ-ω, k-epsilon (Ƙ-ε, dan Reynold stress model (RSM. Pada analisis sensitivitas aliran turbulen di dalam kanal digunakan model mesh hexahedral dengan memilih tiga geometri sel yang masing masing berukuran 0,5 mm; 0,2 mm dan 0,15 mm. Hasil analisis menunjukkan bahwa pada analisis kondisi tunak (steady state, terdapat hasil yang mirip pada model turbulen Ƙ-ε standard dan Ƙ-ω standard. Pengujian terhadap kriteria Dittus Boelter untuk bilangan Nusselt menunjukkan bahwa model Reynold stress model (RSM direkomendasikan. Analisis sensitivitas terhadap geometri mesh antara sel yang berukuran 0,5 mm, 0,2 mm dan 0,15 mm, menunjukkan bahwa geometri sel sebesar 0,5 mm telah mencukupi. Aliran turbulen berkembang penuh telah tercapai pada model LES dan DES, meskipun hanya dalam waktu singkat (3 s, model LES memerlukan waktu komputasi

  7. Measuring the coolness of interactive products: the COOL questionnaire

    DEFF Research Database (Denmark)

    Bruun, Anders; Raptis, Dimitrios; Kjeldskov, Jesper

    2016-01-01

    characteristics. These were used to create an initial pool of question items and 2236 participants were asked to assess 16 mobile devices. By performing exploratory and confirmatory factor analyses, we identified three factors that can measure the perceived inner coolness of interactive products: desirability...... is the COOL questionnaire. We based the creation of the questionnaire on literature suggesting that perceived coolness is decomposed to outer cool (the style of a product) and inner cool (the personality characteristics assigned to it). In this paper, we focused on inner cool, and we identified 11 inner cool...

  8. The Verification of Coupled Neutronics Thermal-Hydraulics Code NODAL3 in the PWR Rod Ejection Benchmark

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2014-01-01

    Full Text Available A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of this work is to determine the accuracy of NODAL3 code in analysing the reactivity initiated accident due to the control rod ejection. The NEACRP PWR rod ejection cases are chosen since many organizations participated in the NEA project using various methods as well as approximations, so that, in addition to the reference solutions, the calculation results of NODAL3 code can also be compared to other codes’ results. The transient parameters to be verified are time of power peak, power peak, final power, final average Doppler temperature, maximum fuel temperature, and final coolant temperature. The results of NODAL3 code agree well with the PHANTHER reference solutions in 1993 and 1997 (revised. Comparison with other validated codes, DYN3D/R and ANCK, shows also a satisfactory agreement.

  9. Comparative study of the contribution of various PWR spacer grid components to hydrodynamic and wall pressure characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Bhattacharjee, Saptarshi, E-mail: saptarshi.bhattacharjee@outlook.com [Alternative Energies and Atomic Energy Commission (CEA) – Cadarache, DEN/DTN/STCP/LHC, 13108 Saint Paul lez Durance Cedex (France); Laboratoire de Mécanique, Modélisation et Procédés Propres (M2P2), UMR7340 CNRS, Aix-Marseille Université, Centrale Marseille, 13451 Marseille Cedex (France); Ricciardi, Guillaume [Alternative Energies and Atomic Energy Commission (CEA) – Cadarache, DEN/DTN/STCP/LHC, 13108 Saint Paul lez Durance Cedex (France); Viazzo, Stéphane [Laboratoire de Mécanique, Modélisation et Procédés Propres (M2P2), UMR7340 CNRS, Aix-Marseille Université, Centrale Marseille, 13451 Marseille Cedex (France)

    2017-06-15

    Highlights: • Complex geometry inside a PWR fuel assembly is simulated using simplified 3D models. • Structured meshes are generated as far as possible. • Fluctuating hydrodynamic and wall pressure field are analyzed using LES. • Comparative studies between square spacer grid, circular spacer grid and mixing vanes are presented. • Simulations are compared with experimental data. - Abstract: Flow-induced vibrations in a pressurized water reactor (PWR) core can cause fretting wear in fuel rods. These vibrations can compromise safety of a nuclear reactor. So, it is necessary to know the random fluctuating forces acting on the rods which cause these vibrations. In this paper, simplified 3D models like square spacer grid, circular spacer grid and symmetric mixing vanes have been used inside an annular pipe. Hydrodynamic and wall pressure characteristics are evaluated using large eddy simulations (LES). Structured meshes are generated as far as possible. Simulations are compared with an experiment. Results show that the grid and vanes have a combined effect: grid accelerates the flow whereas the vanes contribute to the swirl structures. Spectral analysis of the simulations illustrate vortex shedding phenomenon in the wake of spacer grids. This initial study opens up interesting perspectives towards improving the modeling strategy and understanding the complex phenomenon inside a PWR core.

  10. Organ-specific gene expression in maize: The P-wr allele. Final report, August 15, 1993--August 14, 1996

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, T.A.

    1997-06-01

    The ultimate aim of our work is to understand how a regulatory gene produces a specific pattern of gene expression during plant development. Our model is the P-wr gene of maize, which produces a distinctive pattern of pigmentation of maize floral organs. We are investigating this system using a combination of classical genetic and molecular approaches. Mechanisms of organ-specific gene expression are a subject of intense research interest, as it is the operation of these mechanisms during eukaryotic development which determine the characteristics of each organism Allele-specific expression has been characterized in only a few other plant genes. In maize, organ-specific pigmentation regulated by the R, B, and Pl genes is achieved by differential transcription of functionally conserved protein coding sequences. Our studies point to a strikingly different mechanism of organ-specific gene expression, involving post-transcriptional regulation of the regulatory P gene. The novel pigmentation pattern of the P-wr allele is associated with differences in the encoded protein. Furthermore, the P-wr gene itself is present as a unique tandemly amplified structure, which may affect its transcriptional regulation.

  11. Characterization of PWR vessel steel tearing under severe accident condition temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Matheron, Philippe, E-mail: philippe.matheron@cea.fr [CEA, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Chapuliot, Stephane, E-mail: stephane.chapuliot@cea.fr [CEA, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Nicolas, Laetitia, E-mail: laetitia.nicolas@cea.fr [CEA, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Laboratoire de Mecanique des Structures Industrielles Durables, UMR CNRS-EDF 2832, 1 avenue du General de Gaulle, F-92141 Clamart (France); Koundy, Vincent, E-mail: vincent.koundy@irsn.fr [IRSN-DSR, Service d' evaluation des Accidents Graves et des Rejets radioactifs B.P. 17, 92262 Fontenay-aux-Roses Cedex (France); Caroli, Cataldo, E-mail: cataldo.caroli@irsn.fr [IRSN-DSR, Service d' evaluation des Accidents Graves et des Rejets radioactifs B.P. 17, 92262 Fontenay-aux-Roses Cedex (France)

    2012-01-15

    Highlights: Black-Right-Pointing-Pointer We characterized French PWR vessel steel tearing resistance at high temperatures. Black-Right-Pointing-Pointer Tearing tests on Compact Tension (CT) specimens were carried out. Black-Right-Pointing-Pointer The variability of tearing properties with PWR vessels specifications was studied. Black-Right-Pointing-Pointer We propose a tearing criterion (energy parameter Gfr) at high temperatures. - Abstract: In the event of a severe core meltdown accident in a pressurised water reactor (PWR), core material can relocate into the lower head of the vessel resulting in significant thermal and pressure loads being imposed on the vessel. In the event of reactor pressure vessel (RPV) failure there is the possibility of core material being released towards the containment. On the basis of the loading conditions and the temperature distribution, the determination of the mode, timing, and size of lower head failure is of prime importance in the assessment of core melt accidents. This is because they define the initial conditions for ex-vessel events such as core/basemat interactions, fuel/coolant interactions, and direct containment heating. When lower head failure occurs (i) the understanding of the mechanism of lower head creep deformation; (ii) breach stability and its kinetic of propagation leading to the failure; (iii) and developing predictive modelling capabilities to better assess the consequences of ex-vessel processes, are of equal importance. The objective of this paper is to present an original characterization programme of vessel steel tearing properties by carrying out high temperature tearing tests on Compact Tension (CT) specimens. The influence of metallurgical composition on the kinetics of tearing is investigated as previous work on different RPV steels has shown a possible loss of ductility at high temperatures depending on the initial chemical composition of the vessel material. Small changes in the composition can lead

  12. Preliminary Neutronic Study of D2O-cooled High Conversion PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Hikaru Hiruta; Gilles Youinou

    2013-10-01

    This paper presents a preliminary neutronics analysis of tight-pitch D2O-cooled high-conversion PWRs loaded with MOX fuel aiming at high Pu conversion and negative void coefficient. SCALE6.1 has been exclusively utilized for this study. The analyses are performed in two separate parts. The first part of this paper investigates the performance of axial and internal blankets and seeks break-even or near-breeder core even without the presence of radial blankets. The second part of this paper performs sensitivity and uncertainty analyses of integral parameters (keff and void coefficient) for selected systems in order to analyze the characters of this high-conversion PWR from different aspects.

  13. The definition of cool

    Energy Technology Data Exchange (ETDEWEB)

    Nichiporuk, A.

    2005-05-01

    A new air cooling system at Agnico-Eagle's LaRonde mine, located in the Abitibi Region of Quebec is described. The new system serves a mine operating at 7,250 plus feet level. The system is installed at the surface; it utilizes ammonia to cool water, which cools the air. The system consists of four compressors which lower the temperature of the ammonia to minus 2 degrees C. Water, which at this temperature is 14 degrees, and ammonia pass through a plate heat exchanger simultaneously, however, without coming into contact with each other. The heat transfer that occurs causes the water's temperature to drop to 2 degrees C. The total volume of water cooled is 220 litres per second. The system is capable of reducing 636,000 cfm of air from 30 degrees C to 6 degrees C, to which 214,000 cfm of non-cooled air is added. This mixture, which is maintained at approximately 8 degrees C throughout the summer season, is sent underground to the deepest parts of the mine. The system runs from June to September, depending on the weather. In the evenings, when the temperature dips to around four to five degrees C, the water is shut down and side doors are opened to prevent the water from freezing.

  14. Aspects of Household Cooling Technology

    OpenAIRE

    Mrzyglod, Matthias; Holzer, Stefan

    2014-01-01

    Actually available household cooling appliances in the highest efficiency class may consume less then 10W average electrical power. To achieve such power consumptions special challenges for the cooling system had to overcome. The related cooling system design has to consider several effects, which arise from the corresponding low cooling capacity demand, start/stop cycles and additional power consumptions by control accessories. The lecture provides symptomatic aspects of cooling technology, ...

  15. Doppler cooling a microsphere

    CERN Document Server

    Barker, P F

    2010-01-01

    Doppler cooling the center-of-mass motion of an optically levitated microsphere via the velocity dependent scattering force from narrow whispering gallery mode (WGM) resonances is described. Light that is red detuned from the WGM resonance can be used to damp the center-of-mass motion in a process analogous to the Doppler cooling of atoms. Leakage of photons out of the microsphere when the incident field is near resonant with the narrow WGM resonance acts to damp the motion of the sphere. The scattering force is not limited by saturation, but can be controlled by the incident power. Cooling times on the order of seconds are calculated for a 20 micron diameter silica microsphere trapped within optical tweezers, with a Doppler temperature limit in the microKelvin regime.

  16. Natural radiative cooling

    Energy Technology Data Exchange (ETDEWEB)

    Lazzarin, R.

    1979-01-01

    Natural radiative cooling at night was measured based on the surface-radiation spectrum after the heat balance of the surface exposed to the sun had been reradiated. A concept equivalent to the sky temperature and a concept useful for obtaining the net heat flux are discussed. The highest possible equilibrium temperature of the selective surface can be lowered; however, how to apply this practically is not yet known. A simple radiator, completely enclosed by a transparent screen, can produce a significant and inexpensive cooling effect. The results of experiments carried out in an area such as Padua, Italy, where the climate is not suitable for cooling purposes can still be predicted theoretically. The possibility of using the collector for heat collection during the day and as a radiator at night is indicated.

  17. Clean cooling; Saubere Kuehlung

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1998-07-01

    The round hybrid cooling towers which Balcke-Duerr GmbH is currently building for the 550-MW-IGCC-power-station of a refinery project on Sardinia have to meet particularly stringent requirements as seawater is used for cooling. The advantages are: Avoidance of visible plume with minimal energy consumption, optimal plume exit velocity and discharge, greatest possible stability of the plume column, avoidance of interference and recirculation, high operating reliability of the cooling tower. (orig.) [Deutsch] Derzeit werden die Kuehltuerme fuer ein 550-MW-IGCC-Kraftwerk einer Raffinierie auf Sardinien errichtet. Die Anforderungen an die Technik sind hoch, denn gekuehlt wird mit Seewasser. Zum Einsatz kommen Hybridrundkuehltuerme der Balcke-Duerr GmbH, Ratingen. Damit setzt das Unternehmen diesen Typ erstmals ausserhalb von Deutschland ein. (orig.)

  18. Monitoring Cray Cooling Systems

    Energy Technology Data Exchange (ETDEWEB)

    Maxwell, Don E [ORNL; Ezell, Matthew A [ORNL; Becklehimer, Jeff [Cray, Inc.; Donovan, Matthew J [ORNL; Layton, Christopher C [ORNL

    2014-01-01

    While sites generally have systems in place to monitor the health of Cray computers themselves, often the cooling systems are ignored until a computer failure requires investigation into the source of the failure. The Liebert XDP units used to cool the Cray XE/XK models as well as the Cray proprietary cooling system used for the Cray XC30 models provide data useful for health monitoring. Unfortunately, this valuable information is often available only to custom solutions not accessible by a center-wide monitoring system or is simply ignored entirely. In this paper, methods and tools used to harvest the monitoring data available are discussed, and the implementation needed to integrate the data into a center-wide monitoring system at the Oak Ridge National Laboratory is provided.

  19. Experimental study on the heat transfer characteristics of a nuclear reactor containment wall cooled by gravitationally falling water

    Science.gov (United States)

    Pasek, Ari D.; Umar, Efrison; Suwono, Aryadi; Manalu, Reinhard E. E.

    2012-06-01

    Gravitationally falling water cooling is one of mechanism utilized by a modern nuclear Pressurized Water Reactor (PWR) for its Passive Containment Cooling System (PCCS). Since the cooling is closely related to the safety, water film cooling characteristics of the PCCS should be studied. This paper deals with the experimental study of laminar water film cooling on the containment model wall. The influences of water mass flow rate and wall heat rate on the heat transfer characteristic were studied. This research was started with design and assembly of a containment model equipped with the water cooling system, and calibration of all measurement devices. The containment model is a scaled down model of AP 1000 reactor. Below the containment steam is generated using electrical heaters. The steam heated the containment wall, and then the temperatures of the wall in several positions were measure transiently using thermocouples and data acquisition. The containment was then cooled by falling water sprayed from the top of the containment. The experiments were done for various wall heat rate and cooling water flow rate. The objective of the research is to find the temperature profile along the wall before and after the water cooling applied, prediction of the water film characteristic such as means velocity, thickness and their influence to the heat transfer coefficient. The result of the experiments shows that the wall temperatures significantly drop after being sprayed with water. The thickness of water film increases with increasing water flow rate and remained constant with increasing wall heat rate. The heat transfer coefficient decreases as film mass flow rate increase due to the increases of the film thickness which causes the increasing of the thermal resistance. The heat transfer coefficient increases slightly as the wall heat rate increases. The experimental results were then compared with previous theoretical studied.

  20. Research on Cooling Effectiveness in Stepped Slot Film Cooling Vane

    Institute of Scientific and Technical Information of China (English)

    LI Yulong; WU Hong; ZHOU Feng; RONG Chengjun

    2016-01-01

    As one of the most important developments in air cooling technology for hot parts of the aero-engine,film cooling technology has been widely used.Film cooling hole structure exists mainly in areas that have high temperature,uneven cooling effectiveness issues when in actual use.The first stage turbine vanes of the aero-engine consume the largest portion of cooling air,thereby the research on reducing the amount of cooling air has the greatest potential.A new stepped slot film cooling vane with a high cooling effectiveness and a high cooling uniformity was researched initially.Through numerical methods,the affecting factors of the cooling effectiveness of a vane with the stepped slot film cooling structure were researched.This paper focuses on the cooling effectiveness and the pressure loss in different blowing ratio conditions,then the most reasonable and scientific structure parameter can be obtained by analyzing the results.The results show that 1.0 mm is the optimum slot width and 10.0 is the most reasonable blowing ratio.Under this condition,the vane achieved the best cooling result and the highest cooling effectiveness,and also retained a low pressure loss.

  1. Mitigation of stress corrosion cracking in pressurized water reactor (PWR) piping systems using the mechanical stress improvement process (MSIP{sup R)} or underwater laser beam welding

    Energy Technology Data Exchange (ETDEWEB)

    Rick, Grendys; Marc, Piccolino; Cunthia, Pezze [Westinghouse Electric Company, LLC, New York (United States); Badlani, Manu [Nu Vision Engineering, New York (United States)

    2009-04-15

    A current issue facing pressurized water reactors (PWRs) is primary water stress corrosion cracking (PWSCC) of bi metallic welds. PWSCC in a PWR requires the presence of a susceptible material, an aggressive environment and a tensile stress of significant magnitude. Reducing the potential for SCC can be accomplished by eliminating any of these three elements. In the U.S., mitigation of susceptible material in the pressurizer nozzle locations has largely been completed via the structural weld overlay (SWOL) process or NuVision Engineering's Mechanical Stress Improvement Process (MSIP{sup R)}, depending on inspectability. The next most susceptible locations in Westinghouse designed power plants are the Reactor Vessel (RV) hot leg nozzle welds. However, a full SWOL Process for RV nozzles is time consuming and has a high likelihood of in process weld repairs. Therefore, Westinghouse provides two distinctive methods to mitigate susceptible material for the RV nozzle locations depending on nozzle access and utility preference. These methods are the MSIP and the Underwater Laser Beam Welding (ULBW) process. MSIP applies a load to the outside diameter of the pipe adjacent to the weld, imposing plastic strains during compression that are not reversed after unloading, thus eliminating the tensile stress component of SCC. Recently, Westinghouse and NuVision successfully applied MSIP on all eight RV nozzles at the Salem Unit 1 power plant. Another option to mitigate SCC in RV nozzles is to place a barrier between the susceptible material and the aggressive environment. The ULBW process applies a weld inlay onto the inside pipe diameter. The deposited weld metal (Alloy 52M) is resistant to PWSCC and acts as a barrier to prevent primary water from contacting the susceptible material. This paper provides information on the approval and acceptance bases for MSIP, its recent application on RV nozzles and an update on ULBW development.

  2. Anomalous law of cooling.

    Science.gov (United States)

    Lapas, Luciano C; Ferreira, Rogelma M S; Rubí, J Miguel; Oliveira, Fernando A

    2015-03-14

    We analyze the temperature relaxation phenomena of systems in contact with a thermal reservoir that undergoes a non-Markovian diffusion process. From a generalized Langevin equation, we show that the temperature is governed by a law of cooling of the Newton's law type in which the relaxation time depends on the velocity autocorrelation and is then characterized by the memory function. The analysis of the temperature decay reveals the existence of an anomalous cooling in which the temperature may oscillate. Despite this anomalous behavior, we show that the variation of entropy remains always positive in accordance with the second law of thermodynamics.

  3. Anomalous law of cooling

    Science.gov (United States)

    Lapas, Luciano C.; Ferreira, Rogelma M. S.; Rubí, J. Miguel; Oliveira, Fernando A.

    2015-03-01

    We analyze the temperature relaxation phenomena of systems in contact with a thermal reservoir that undergoes a non-Markovian diffusion process. From a generalized Langevin equation, we show that the temperature is governed by a law of cooling of the Newton's law type in which the relaxation time depends on the velocity autocorrelation and is then characterized by the memory function. The analysis of the temperature decay reveals the existence of an anomalous cooling in which the temperature may oscillate. Despite this anomalous behavior, we show that the variation of entropy remains always positive in accordance with the second law of thermodynamics.

  4. Superconductor rotor cooling system

    Science.gov (United States)

    Gamble, Bruce B.; Sidi-Yekhlef, Ahmed; Schwall, Robert E.; Driscoll, David I.; Shoykhet, Boris A.

    2002-01-01

    A system for cooling a superconductor device includes a cryocooler located in a stationary reference frame and a closed circulation system external to the cryocooler. The closed circulation system interfaces the stationary reference frame with a rotating reference frame in which the superconductor device is located. A method of cooling a superconductor device includes locating a cryocooler in a stationary reference frame, and transferring heat from a superconductor device located in a rotating reference frame to the cryocooler through a closed circulation system external to the cryocooler. The closed circulation system interfaces the stationary reference frame with the rotating reference frame.

  5. A Cool Emperor Penguin

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    哇,这只帝企鹅的胸前居然有个心形!It’s cool!(乐天:没错,它的胸前少了这幺撮毛,应该会感觉挺凉快的)cool在这里可不是“凉快”的意思,而是“酷.帅气”的意思。我们《英语大王》的英文名字就叫English Cool Kids哦!(乐天拿出一副墨镜戴上:

  6. Rapid cooled lens cell

    Science.gov (United States)

    Stubbs, David M.; Hsu, Ike C.

    1991-12-01

    This paper describes the optomechanical design, thermal analysis, fabrication, and test evaluation processes followed in developing a rapid cooled, infrared lens cell. Thermal analysis was the key engineering discipline exercised in the design phase. The effect of thermal stress on the lens, induced by rapid cooling of the lens cell, was investigated. Features of this lens cell that minimized the thermal stress will be discussed in a dedicated section. The results of thermal analysis on the selected lens cell design and the selection of the flow channel design in the heat exchanger will be discussed. Throughout the paper engineering drawings, illustrations, analytical results, and photographs of actual hardware are presented.

  7. Gas cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1972-06-01

    Although most of the development work on fast breeder reactors has been devoted to the use of liquid metal cooling, interest has been expressed for a number of years in alternative breeder concepts using other coolants. One of a number of concepts in which interest has been retained is the Gas-Cooled Fast Reactor (GCFR). As presently envisioned, it would operate on the uranium-plutonium mixed oxide fuel cycle, similar to that used in the Liquid Metal Fast Breeder Reactor (LMFBR), and would use helium gas as the coolant.

  8. Quantum thermodynamic cooling cycle

    CERN Document Server

    Palao, J P; Gordon, J M; Palao, Jose P.; Kosloff, Ronnie; Gordon, Jeffrey M.

    2001-01-01

    The quantum-mechanical and thermodynamic properties of a 3-level molecular cooling cycle are derived. An inadequacy of earlier models is rectified in accounting for the spontaneous emission and absorption associated with the coupling to the coherent driving field via an environmental reservoir. This additional coupling need not be dissipative, and can provide a thermal driving force - the quantum analog of classical absorption chillers. The dependence of the maximum attainable cooling rate on temperature, at ultra-low temperatures, is determined and shown to respect the recently-established fundamental bound based on the second and third laws of thermodynamics.

  9. Validation of the scale system for PWR spent fuel isotopic composition analyses

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.; Bowman, S.M.; Parks, C.V. [Oak Ridge National Lab., TN (United States); Brady, M.C. [Sandia National Laboratories, Las Vegas, NV (United States)

    1995-03-01

    The validity of the computation of pressurized-water-reactor (PWR) spent fuel isotopic composition by the SCALE system depletion analysis was assessed using data presented in the report. Radiochemical measurements and SCALE/SAS2H computations of depleted fuel isotopics were compared with 19 benchmark-problem samples from Calvert Cliffs Unit 1, H. B. Robinson Unit 2, and Obrigheim PWRs. Even though not exhaustive in scope, the validation included comparison of predicted and measured concentrations for 14 actinides and 37 fission and activation products. The basic method by which the SAS2H control module applies the neutron transport treatment and point-depletion methods of SCALE functional modules (XSDRNPM-S, NITAWL-II, BONAMI, and ORIGEN-S) is described in the report. Also, the reactor fuel design data, the operating histories, and the isotopic measurements for all cases are included in detail. The underlying radiochemical assays were conducted by the Materials Characterization. Center at Pacific Northwest Laboratory as part of the Approved Testing Material program and by four different laboratories in Europe on samples processed at the Karlsruhe Reprocessing Plant.

  10. VOF Calculations of Countercurrent Gas-Liquid Flow in a PWR Hot Leg

    Directory of Open Access Journals (Sweden)

    M. Murase

    2012-01-01

    Full Text Available We improved the computational grid and schemes in the VOF (volume of fluid method with the standard − turbulent model in our previous study to evaluate CCFL (countercurrent flow limitation characteristics in a full-scale PWR hot leg (750 mm diameter, and the calculated CCFL characteristics agreed well with the UPTF data at 1.5 MPa. In this paper, therefore, to evaluate applicability of the VOF method to different fluid properties and a different scale, we did numerical simulations for full-scale air-water conditions and the 1/15-scale air-water tests (50 mm diameter, respectively. The results calculated for full-scale conditions agreed well with CCFL data and showed that CCFL characteristics in the Wallis diagram were mitigated under 1.5 MPa steam-water conditions comparing with air-water flows. However, the results calculated for the 1/15-scale air-water tests greatly underestimated the falling water flow rates in calculations with the standard − turbulent model, but agreed well with the CCFL data in calculations with a laminar flow model. This indicated that suitable calculation models and conditions should be selected to get good agreement with data for each scale.

  11. Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises

    Energy Technology Data Exchange (ETDEWEB)

    El Bakkari, B. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco)], E-mail: bakkari@gmail.com; El Bardouni, T.; Merroun, O.; El Younoussi, Ch.; Boulaich, Y. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco); Chakir, E. [EPTN-LPMR, Faculty of Sciences Kenitra (Morocco)

    2009-05-15

    The aim of this study is to evaluate the capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5, through the use of the MCNP tally information. BUCAL1 uses the fourth order Runge Kutta method with the predictor-corrector approach as the integration method to determine the fuel composition at a desired burnup step. Validation of BUCAL1 was done by code vs. code comparison. Results of two different kinds of codes are employed. The first one is CASMO-4, a deterministic multi-group two-dimensional transport code. The second kind is MCODE and MOCUP, a link MCNP-ORIGEN codes. These codes use different burnup algorithms to solve the depletion equations system. Eigenvalue and isotope concentrations were compared for two PWR uranium and thorium benchmark exercises at cold (300 K) and hot (900 K) conditions, respectively. The eigenvalue comparison between BUCAL1 and the aforementioned two kinds of codes shows a good prediction of the systems'k-inf values during the entire burnup history, and the maximum difference is within 2%. The differences between the BUCAL1 isotope concentrations and the predictions of CASMO-4, MCODE and MOCUP are generally better, and only for a few sets of isotopes these differences exceed 10%.

  12. Test requirements for the integral effect test to simulate Korean PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Song, Chul Hwa; Park, C. K.; Lee, S. J.; Kwon, T. S.; Yun, B. J.; Chung, M. K

    2001-02-01

    In this report, the test requirements are described for the design of the integral effect test facility to simulate Korean PWR plants. Since the integral effect test facility should be designed so as to simulate various thermal hydraulic phenomena, as closely as possible, to be occurred in real plants during operation or anticipated transients, the design and operational characteristics of the reference plants (Korean Standard Nuclear Plant and Korean Next Generation Reactor)were analyzed in order to draw major components, systems, and functions to be satisfied or simulated in the test facility. The test matrix is set up by considering major safety concerns of interest and the test objectives to confirm and enhance the safety of the plants. And the analysis and prioritization of the test matrix leads to the general design requirements of the test facility. Based on the general design requirements, the design criteria is set up for the basic and detailed design of the test facility. And finally it is drawn the design requirements specific to the fluid system and measurement system of the test facility. The test requirements in this report will be used as a guideline to the scaling analysis and basic design of the test facility. The test matrix specified in this report can be modified in the stage of main testing by considering the needs of experiments and circumstances at that time.

  13. Analysis of measured and calculated counterpart test data in PWR and VVER 1000 simulators

    Directory of Open Access Journals (Sweden)

    d’Auria Francesco

    2005-01-01

    Full Text Available This paper presents an over view of the "scaling strategy", in particular the role played by the counter part test methodology. The recent studies dealing with a scaling analysis in light water reactor with special regard to the VVER 1000 Russian reactor type are presented to demonstrate the phenomena important for scaling. The adopted scaling approach is based on the selection of a few characteristic parameters chosen by taking into account their relevance in the behavior of the transient. The adopted computer code used is RELAP5/Mod3.3 and its accuracy has been demonstrated by qualitative and quantitative evaluation. Comparing experimental data, it was found that the investigated facilities showed similar behavior concerning the time trends, and that the same thermal hydraulic phenomena on a qualitative level could be predicted. The main results are: PSB and LOBI main parameters have similar trends. This fact is the confirmation of the validity of the adopted scaling approach and it shows that PWR and VVER reactor type behavior is very similar. No new phenomena occurred during the counter part test, despite the fact that the two facilities had a different lay out, and the already known phenomena were predicted correctly by the code. The code capability and accuracy are scale-independent. Both character is tics are necessary to permit the full scale calculation with the aim of nuclear power plant behavior prediction. .

  14. Computer simulation of Angra-2 PWR nuclear reactor core using MCNPX code

    Energy Technology Data Exchange (ETDEWEB)

    Medeiros, Marcos P.C. de; Rebello, Wilson F., E-mail: eng.cavaliere@ime.eb.br, E-mail: rebello@ime.eb.br [Instituto Militar de Engenharia - Secao de Engenharia Nuclear, Rio de Janeiro, RJ (Brazil); Oliveira, Claudio L. [Universidade Gama Filho, Departamento de Matematica, Rio de Janeiro, RJ (Brazil); Vellozo, Sergio O., E-mail: vellozo@cbpf.br [Centro Tecnologico do Exercito. Divisao de Defesa Quimica, Biologica e Nuclear, Rio de Janeiro, RJ (Brazil); Silva, Ademir X. da, E-mail: ademir@nuclear.ufrj.br [Coordenacao dos Programas de Pos Gaduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    In this work the MCNPX (Monte Carlo N-Particle Transport Code) code was used to develop a computerized model of the core of Angra 2 PWR (Pressurized Water Reactor) nuclear reactor. The model was created without any kind of homogenization, but using real geometric information and material composition of that reactor, obtained from the FSAR (Final Safety Analysis Report). The model is still being improved and the version presented in this work is validated by comparing values calculated by MCNPX with results calculated by others means and presented on FSAR. This paper shows the results already obtained to K{sub eff} and K{infinity}, general parameters of the core, considering the reactor operating under stationary conditions of initial testing and operation. Other stationary operation conditions have been simulated and, in all tested cases, there was a close agreement between values calculated computationally through this model and data presented on the FSAR, which were obtained by other codes. This model is expected to become a valuable tool for many future applications. (author)

  15. Study of chemical additives in the cementation of radioactive waste of PWR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Vieira, Vanessa Mota; Tello, Cledola Cassia Oliveira de, E-mail: vanessamotavieira@gmail.com, E-mail: tellocc@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2012-03-15

    In this research it has been studied the effects of chemical admixtures in the cementation process of radioactive wastes. These additives are used to improve the properties of waste cementation process, both of the paste and of the solidified product. However there are a large variety of these materials that are frequently changed or taken out of the market. Then it is essential to know the commercially available materials and their effects. The tests were carried out with a solution simulating the evaporator concentrate waste coming from PWR nuclear reactors. It was cemented using two formulations, A and B, incorporating higher or lower amount of waste, respectively. It was added chemical admixtures from two manufacturers (S and H), which were: accelerators, set retarders and superplasticizers. The experiments were organized by a factorial design 23. The measured parameters were: the viscosity, the setting time, the paste and product density and the compressive strength. The parameter evaluated in this study was the compressive strength at age of 28 days, is considered essential security issues relating to the handling, transport and storage of cemented waste product. The results showed that the addition of accelerators improved the compressive strength of the cemented products. (author)

  16. Numerical Simulation of Size Effects on Countercurrent Flow Limitation in PWR Hot Leg Models

    Directory of Open Access Journals (Sweden)

    I. Kinoshita

    2012-01-01

    Full Text Available We have previously done numerical simulations using the two-fluid model implemented in the CFD software FLUENT6.3.26 to investigate effects of shape of a flow channel and its size on CCFL (countercurrent flow limitation characteristics in PWR hot leg models. We confirmed that CCFL characteristics in the hot leg could be well correlated with the Wallis parameters in the diameter range of 0.05 m≤D≤0.75 m. In the present study, we did numerical simulations using the two-fluid model for the air-water tests with D=0.0254 m to determine why CCFL characteristics for D=0.0254 m were severer compared with those in the range, 0.05 m≤D≤0.75 m. The predicted CCFL characteristics agreed with the data for D=0.0254 m and indicated that the CCFL difference between D=0.0254 m and 0.05 mm≤D≤0.75 mm was caused by the size effect and not by other factors.

  17. Risk-Informed External Hazards Analysis for Seismic and Flooding Phenomena for a Generic PWR

    Energy Technology Data Exchange (ETDEWEB)

    Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Prescott, Steve [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ma, Zhegang [Idaho National Lab. (INL), Idaho Falls, ID (United States); Spears, Bob [Idaho National Lab. (INL), Idaho Falls, ID (United States); Szilard, Ronaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Coleman, Justin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kosbab, Ben [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-26

    This report describes the activities performed during the FY2017 for the US-DOE Light Water Reactor Sustainability Risk-Informed Safety Margin Characterization (LWRS-RISMC), Industry Application #2. The scope of Industry Application #2 is to deliver a risk-informed external hazards safety analysis for a representative nuclear power plant. Following the advancements occurred during the previous FYs (toolkits identification, models development), FY2017 focused on: increasing the level of realism of the analysis; improving the tools and the coupling methodologies. In particular the following objectives were achieved: calculation of buildings pounding and their effects on components seismic fragility; development of a SAPHIRE code PRA models for 3-loops Westinghouse PWR; set-up of a methodology for performing static-dynamic PRA coupling between SAPHIRE and EMRALD codes; coupling RELAP5-3D/RAVEN for performing Best-Estimate Plus Uncertainty analysis and automatic limit surface search; and execute sample calculations for demonstrating the capabilities of the toolkit in performing a risk-informed external hazards safety analyses.

  18. PWR composite materials use. A particular case of safety-related service water pipes

    Energy Technology Data Exchange (ETDEWEB)

    Pays, M.F.; Le Courtois, T

    1997-11-01

    This paper shows the present and future uses of composite materials in French nuclear and fossil-fuel power plants. Electricite de France has decided to install composite materials in service water piping in its future nuclear power plant (PWR) at Civaux (West of France) and for the firs time in France, in safety-related applications. A wide range of studies has been performed about the durability, the control and damage mechanisms of those materials under service conditions among an ongoing Research and Development project. The main results are presented under the following headlines: selection of basic materials and manufacturing processes; aging processes (mechanical behavior during `lifetime`); design rules; non destructive examination during manufacturing process and during operation. The studies have been focused on epoxy pipings. The importance of strong quality insurance policy requirements are outlined. A study of the use of composite pipes in power plants (hydraulic, fossil fuel, and nuclear) in France and around the world (USA, Japan, Western Europe) are presented whether it be safety related or non safety-related applications. The different technical solutions for materials and manufacturing processes are presented and an economic comparison is made between steel and composite pipes. (author) 2 refs.

  19. Fuel failure and fission gas release in high burnup PWR fuels under RIA conditions

    Science.gov (United States)

    Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide; Ishijima, Kiyomi

    1997-09-01

    To study the fuel behavior and to evaluate the fuel enthalpy threshold of fuel rod failure under reactivity initiated accident (RIA) conditions, a series of experiments using pulse irradiation capability of the Nuclear Safety Research Reactor (NSRR) has been performed. During the experiments with 50 MWd/kg U PWR fuel rods (HBO test series; an acronym for high burnup fuels irradiated in Ohi unit 1 reactor), significant cladding failure occurred. The energy deposition level at the instant of the fuel failure in the test is 60 cal/g fuel, and is considerably lower than those expected and pre-evaluated. The result suggests that mechanical interaction between the fuel pellets and the cladding tube with decreased integrity due to hydrogen embrittlement causes fuel failure at the low energy deposition level. After the pulse irradiation, the fuel pellets were found as fragmented debris in the coolant water, and most of these were finely fragmented. This paper describes several key observations in the NSRR experiments, which include cladding failure at the lower enthalpy level, possible post-failure events and large fission gas release.

  20. Evaluation of Fuel Performance Uncertainty in a PWR HFP RIA Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Joosuk; Woo, Swengwoong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-10-15

    Sensitivity and combined uncertainty studies based on the various kinds of uncertainty sources have been carried out in a PWR hot full power (HFP) condition. - Cladding inner diameter, fuel thermal conductivity, fuel thermal expansion and peak power have induced a significant impact to the fuel enthalpy and temperature. - Cladding hoop strain was strongly affected by the uncertainty parameters of cladding inner diameter, fuel thermal expansion, EPRI-1 CHF and peak power. - Above results are valid in the given analysis condition in this paper. Thereby, the analysis conditions, for example the peak linear heat rate before RIA or peak power and FWHM etc, are changed the results will be changed also. Approved analysis methodology for licensing application in the safety analysis of reactivity initiated accident (RIA) in Korea is based on a conservative approach. But newly introduced safety criteria, described in section 4.2 of NUREG-0800, tend to reduce the margins or depending on the reactor types rod failure is predicted due to the pellet-to-cladding mechanical interaction (PCMI) criteria. Thereby, licensee is trying to improve the margins by utilizing a less conservative approach.

  1. Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Trombetta, Debora M.; Su, Jian, E-mail: dtrombetta@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Chirayath, Sunil S., E-mail: sunilsc@tamu.edu [Department of Nuclear Engineering and Nuclear Security Science and Policy Institute, Texas A and M University, TX (United States)

    2015-07-01

    Computational models of reactors are increasingly used to predict nuclear reactor physics parameters responsible for reactivity changes which could lead to accidents and losses. In this work, preliminary results for criticality coefficient calculation using the Monte Carlo transport code MCNPX were presented for a small PWR. The computational modeling developed consists of the core with fuel elements, radial reflectors, and control rods inside a pressure vessel. Three different geometries were simulated, a single fuel pin, a fuel assembly and the core, with the aim to compare the criticality coefficients among themselves.The criticality coefficients calculated were: Doppler Temperature Coefficient, Coolant Temperature Coefficient, Coolant Void Coefficient, Power Coefficient, and Control Rod Worth. The coefficient values calculated by the MCNP code were compared with literature results, showing good agreement with reference data, which validate the computational model developed and allow it to be used to perform more complex studies. Criticality Coefficient values for the three simulations done had little discrepancy for almost all coefficients investigated, the only exception was the Power Coefficient. Preliminary results presented show that simple modelling as a fuel assembly can describe changes at almost all the criticality coefficients, avoiding the need of a complex core simulation. (author)

  2. Fatigue-crack growth behavior of Type 347 stainless steels under simulated PWR water conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seokmin; Min, Ki-Deuk; Yoon, Ji-Hyun; Kim, Min-Chul; Lee, Bong-Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Fatigue crack growth rate (FCGR) curve of stainless steel exists in ASME code section XI, but it is still not considering the environmental effects. The longer time nuclear power plant is operated, the more the environmental degradation issues of materials pop up. There are some researches on fatigue crack growth rate of S304 and S316, but researches of FCGR of S347 used in Korea nuclear power plant are insufficient. In this study, the FCGR of S347 stainless steel was evaluated in the PWR high temperature water conditions. The FCGRs of S347 stainless steel under pressurized-water conditions were measured by using compact-tension (CT) specimens at different levels of dissolved oxygen (DO) and frequency. 1. FCGRs of SS347 were slower than that in ASME XI and environmental effect did not occur when frequency was higher than 1Hz. 2. Fatigue crack growth is accelerated by corrosion fatigue and it is more severe when frequency is slower than 0.1Hz. 3. Increase of crack tip opening time increased corrosion fatigue and it deteriorated environmental fatigue properties.

  3. Studies of residual stress measurement and analysis techniques for a PWR dissimilar weld joint

    Energy Technology Data Exchange (ETDEWEB)

    Ogawa, Naoki, E-mail: naoki2_ogawa@mhi.co.jp [Mitsubishi Heavy Industries, Ltd., 2-1-1, Shinhama, Arai-cho, Takasago 676-8686 (Japan); Muroya, Itaru; Iwamoto, Youichi; Ohta, Takahiro; Ochi, Mayumi; Hojo, Kiminobu [Mitsubishi Heavy Industries, Ltd., 2-1-1, Shinhama, Arai-cho, Takasago 676-8686 (Japan); Ogawa, Kazuo [Japan Nuclear Energy Safety Organization, 3-17-1, Toranomon, Minato-ku, Tokyo 105-0001 (Japan)

    2012-02-15

    For evaluation of the PWSCC crack propagation behavior, a test model was produced using the same fabrication process of Japanese PWR plants and the stress distribution change was measured during a fabrication process such as a hydrostatic test, welding a main coolant pipe to the stainless steel safe end and an operation condition test. For confirmation of validity of the numerical estimation method of the stress distribution, FE analysis was performed to calculate the stress distributions for each fabrication process. From the validation procedure, a standard residual stress evaluation method was established. Furthermore for consideration of characteristics of PWSCC's propagation behavior of the dissimilar welding joint of the safe end nozzles, the influence coefficients at the deepest point for the stress intensity factors of axial cracks with large aspect ratio a/c (crack depth/half of surface crack length) was prepared. The crack shape was assumed a rectangular shape and the stress intensity factors at the deepest point of the crack were calculated with change of crack depth using FE analysis. By using these stress distribution and influence coefficients, a behavior of a PWSCC crack propagation at the safe end nozzles can be estimated easily and rationally.

  4. Conceptual Core Analysis of Long Life PWR Utilizing Thorium-Uranium Fuel Cycle

    Science.gov (United States)

    Rouf; Su'ud, Zaki

    2016-08-01

    Conceptual core analysis of long life PWR utilizing thorium-uranium based fuel has conducted. The purpose of this study is to evaluate neutronic behavior of reactor core using combined thorium and enriched uranium fuel. Based on this fuel composition, reactor core have higher conversion ratio rather than conventional fuel which could give longer operation length. This simulation performed using SRAC Code System based on library SRACLIB-JDL32. The calculation carried out for (Th-U)O2 and (Th-U)C fuel with uranium composition 30 - 40% and gadolinium (Gd2O3) as burnable poison 0,0125%. The fuel composition adjusted to obtain burn up length 10 - 15 years under thermal power 600 - 1000 MWt. The key properties such as uranium enrichment, fuel volume fraction, percentage of uranium are evaluated. Core calculation on this study adopted R-Z geometry divided by 3 region, each region have different uranium enrichment. The result show multiplication factor every burn up step for 15 years operation length, power distribution behavior, power peaking factor, and conversion ratio. The optimum core design achieved when thermal power 600 MWt, percentage of uranium 35%, U-235 enrichment 11 - 13%, with 14 years operation length, axial and radial power peaking factor about 1.5 and 1.2 respectively.

  5. Performance of monosphere new gel type ion exchange resins for condensate polisher at PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Nakanishi, S.; Nakamura, M.; Asou, K. [Kansai Electric Power Co., Inc., Osaka (Japan); Izumi, T.; Deguchi, T.; Ino, T.; Hagiwara, M.

    1998-12-31

    There are two kinds of ion exchange resins of gel type and porous one which are used as condensate polisher in LWR nuclear power plants. In order to estimate the performance of these resins on the condensate polisher at the secondary cycle of Japanese PWR plants, a column test was performed setting the column test device in Ohi power station unit 1 of the Kansai Electric Power Co., Inc. and the variations of the resin properties and the samples at the end of column were analyzed. The column test showed that the cross-linking degree of the new gel resins used was lower than those of porous ones. The new resins captured larger amounts of Matrix-Diffused Crud than the conventional cation resins before regeneration but not after that. Whereas the surface adsorbed crud was less captured by the new resins than conventional anion resins. However, there were little differences among these resins in respects of rinsing characteristics, sphericity, water quality, break through capacity, etc. At the condensate polisher in the secondary system it was confirmed that new gel resins had almost the same performance as one of the conventional ones and could be applied to the actual plant. (M.N.)

  6. MELCOR 1.8.2 assessment: Surry PWR TMLB` (with a DCH study)

    Energy Technology Data Exchange (ETDEWEB)

    Kmetyk, L.N.; Cole, R.K. Jr.; Smith, R.C.; Summers, R.M.; Thompson, S.L.

    1994-02-01

    MELCOR is a fully integrated, engineering-level computer code, being developed at Sandia National Laboratories for the USNRC. This code models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As part of an ongoing assessment program, the MELCOR computer code has been used to analyze a station blackout transient in Surry, a three-loop Westinghouse PWR. Basecase results obtained with MELCOR 1.8.2 are presented, and compared to earlier results for the same transient calculated using MELCOR 1.8.1. The effects of new models added in MELCOR 1.8.2 (in particular, hydrodynamic interfacial momentum exchange, core debris radial relocation and core material eutectics, CORSOR-Booth fission product release, high-pressure melt ejection and direct containment heating) are investigated individually in sensitivity studies. The progress in reducing numeric effects in MELCOR 1.8.2, compared to MELCOR 1.8.1, is evaluated in both machine-dependency and time-step studies; some remaining sources of numeric dependencies (valve cycling, material relocation and hydrogen burn) are identified.

  7. Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR

    Directory of Open Access Journals (Sweden)

    Brovchenko Mariya

    2017-01-01

    Full Text Available The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR. The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

  8. Measure Guideline: Ventilation Cooling

    Energy Technology Data Exchange (ETDEWEB)

    Springer, D. [Alliance for Residential Building Innovation (ARBI), David, CA (United States); Dakin, B. [Alliance for Residential Building Innovation (ARBI), David, CA (United States); German, A. [Alliance for Residential Building Innovation (ARBI), David, CA (United States)

    2012-04-01

    The purpose of this measure guideline is to provide information on a cost-effective solution for reducing cooling system energy and demand in homes located in hot-dry and cold-dry climates. This guideline provides a prescriptive approach that outlines qualification criteria, selection considerations, and design and installation procedures.

  9. Cooling of Neutron Stars

    Directory of Open Access Journals (Sweden)

    Grigorian H.

    2010-10-01

    Full Text Available We introduce the theoretical basis for modeling the cooling evolution of compact stars starting from Boltzmann equations in curved space-time. We open a discussion on observational verification of different neutron star models by consistent statistics. Particular interest has the question of existence of quark matter deep inside of compact object, which has to have a specific influence on the cooling history of the star. Besides of consideration of several constraints and features of cooling evolution, which are susceptible of being critical for internal structure of hot compact stars we have introduced a method of extraction of the mass distribution of the neutron stars from temperature and age data. The resulting mass distribution has been compared with the one suggested by supernove simulations. This method can be considered as an additional checking tool for the consistency of theoretical modeling of neutron stars. We conclude that the cooling data allowed existence of neutron stars with quark cores even with one-flavor quark matter.

  10. Elementary stochastic cooling

    Energy Technology Data Exchange (ETDEWEB)

    Tollestrup, A.V.; Dugan, G

    1983-12-01

    Major headings in this review include: proton sources; antiproton production; antiproton sources and Liouville, the role of the Debuncher; transverse stochastic cooling, time domain; the accumulator; frequency domain; pickups and kickers; Fokker-Planck equation; calculation of constants in the Fokker-Planck equation; and beam feedback. (GHT)

  11. ELECTRON COOLING FOR RHIC.

    Energy Technology Data Exchange (ETDEWEB)

    BEN-ZVI,I.

    2001-05-13

    The Accelerator Collider Department (CAD) at Brookhaven National Laboratory is operating the Relativistic Heavy Ion Collider (RHIC), which includes the dual-ring, 3.834 km circumference superconducting collider and the venerable AGS as the last part of the RHIC injection chain. CAD is planning on a luminosity upgrade of the machine under the designation RHIC II. One important component of the RHIC II upgrade is electron cooling of RHIC gold ion beams. For this purpose, BNL and the Budker Institute of Nuclear Physics in Novosibirsk entered into a collaboration aimed initially at the development of the electron cooling conceptual design, resolution of technical issues, and finally extend the collaboration towards the construction and commissioning of the cooler. Many of the results presented in this paper are derived from the Electron Cooling for RHIC Design Report [1], produced by the, BINP team within the framework of this collaboration. BNL is also collaborating with Fermi National Laboratory, Thomas Jefferson National Accelerator Facility and the University of Indiana on various aspects of electron cooling.

  12. Strategy for the Operation of Cooling Towers with variable Speed Fans

    CERN Document Server

    Iñigo-Golfín, J

    2001-01-01

    Within the SPS Cooling Water Project at CERN aimed at the reduction of water consumption, this primary open cooling loop will be closed and all the primary cooling circuit components will be upgraded to the new required duty and brought to the necessary safety and operability standards. In particular the tower fans will be fitted with variable frequency drives to replace the existing two speed motors. This paper presents a study to optimize the operation of SPS cooling towers taking into account outdoor conditions (wet and dry bulb temperatures) and the entirety of the primary circuit in which they will operate.

  13. Analysis of the containment of a compact reactor PWR submitted to loss of coolant accident; Analise da contencao de um reator PWR compacto submetido a acidente de perda de refrigerante

    Energy Technology Data Exchange (ETDEWEB)

    Dutra, Alexandre de Souza; Belchior Junior, Antonio; Guimaraes, Leonam dos Santos [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)

    2000-07-01

    In the present paper analyses were done with the computer code RELAP5/MOD2 for rising the process conditions of the containment of a compact reactor PWR of low potency, submitted to Loss of Coolant Accidents (LOCA). The main results obtained were the behavior of maximum conditions of pressure as a function of the available containment free volume. It was also studied the problem of containment sub-compartmentation, that is to say, the possibility of the rupture to happen in restricted spaces generating high sub-compartment peak pressure and, consequently, high strains on the internal structures. (author)

  14. The safety analysis and thermohydraulic methodologies for the power updating analyses in Spanish PWR plants; Methodologias de diseno termohidraulico y de analisis de seguridad en los aumentos de potencia de centrales PWR

    Energy Technology Data Exchange (ETDEWEB)

    Salesa, F.

    2014-02-01

    This article describes the Safety Analysis and Thermohydraulic methodologies used by ENUSA for the Power Updating analyses in Spanish PWR plants of Westinghouse design: Design tools have been developed over the first cycles resulting new correlations of DNB, fitted to the new fuel assemblies, new DNBR calculation methodology and other improvements in the design areas. Using these methodologies, the available margins between design and limit values are wider. These new margins have allowed to accomplish the design criteria under the new power updating operational conditions. (Author)

  15. Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes; Calculo de fuerzas laterales hidraulicas en elementos combustibles tipo PWR con codigos de dinamica de fluidos coputacional

    Energy Technology Data Exchange (ETDEWEB)

    Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

    2016-08-01

    To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

  16. Optimization of the distribution of bars with gadolinium oxide in reactor fuel elements PWR; Optimizacion de la distribucion de barras con oxido de gadolinio en elementos combustibles para reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Melgar Santa Cecilia, P. A.; Velazquez, J.; Ahnert Iglesias, C.

    2014-07-01

    In the schemes of low leakage, currently used in the majority of PWR reactors, it makes use of absorbent consumables for the effective control of the factors of peak, the critical concentration of initial boron and the moderator temperature coefficient. One of the most used absorbing is the oxide of gadolinium, which is integrated within the fuel pickup. Occurs a process of optimization of fuel elements with oxide of gadolinium, which allows for a smaller number of configurations with a low peak factor for bar. (Author)

  17. Sorption cooling: a valid extension to passive cooling

    NARCIS (Netherlands)

    Doornink, D.J.; Burger, J.F.; Brake, ter H.J.M.

    2008-01-01

    Passive cooling has shown to be a very dependable cryogenic cooling method for space missions. Several missions employ passive radiators to cool down their delicate sensor systems for many years, without consuming power, without exporting vibrations or producing electromagnetic interference. So for

  18. Comments on Ionization Cooling Channel Characteristics

    CERN Document Server

    Neuffer, David

    2013-01-01

    Ionization cooling channels with a wide variety of characteristics and cooling properties are being developed. These channels can produce cooling performances that are largely consistent with the ionization cooling theory developed previously. In this paper we review ionization cooling theory, discuss its application to presently developing cooling channels, and discuss criteria for optimizing cooling.

  19. Comments on Ionization Cooling Channel Characteristics

    OpenAIRE

    Neuffer, David

    2013-01-01

    Ionization cooling channels with a wide variety of characteristics and cooling properties are being developed. These channels can produce cooling performances that are largely consistent with the ionization cooling theory developed previously. In this paper we review ionization cooling theory, discuss its application to presently developing cooling channels, and discuss criteria for optimizing cooling.

  20. STOCHASTIC COOLING FOR BUNCHED BEAMS.

    Energy Technology Data Exchange (ETDEWEB)

    BLASKIEWICZ, M.

    2005-05-16

    Problems associated with bunched beam stochastic cooling are reviewed. A longitudinal stochastic cooling system for RHIC is under construction and has been partially commissioned. The state of the system and future plans are discussed.

  1. ALP hints from cooling anomalies

    CERN Document Server

    Giannotti, Maurizio

    2015-01-01

    We review the current status of the anomalies in stellar cooling and argue that, among the new physics candidates, an axion-like particle would represent the best option to account for the hinted additional cooling.

  2. Cooling devices in laser therapy

    Directory of Open Access Journals (Sweden)

    Anupam Das

    2016-01-01

    Full Text Available Cooling devices and methods are now integrated into most laser systems, with a view to protecting the epidermis, reducing pain and erythema and improving the efficacy of laser. On the basis of method employed, it can be divided into contact cooling and non-contact cooling. With respect to timing of irradiation of laser, the nomenclatures include pre-cooling, parallel cooling and post-cooling. The choice of the cooling device is dictated by the laser device, the physician′s personal choice with respect to user-friendliness, comfort of the patient, the price and maintenance costs of the device. We hereby briefly review the various techniques of cooling, employed in laser practice.

  3. Optical cavity cooling of mechanical modes of a semiconductor nanomembrane

    DEFF Research Database (Denmark)

    Usami, Koji; Naesby, A.; Bagci, Tolga

    2012-01-01

    . The resultant photo-induced rigidity is large and a mode temperature cooled from room temperature down to 4 K is realized with 50 μW of light and a cavity finesse of just 10. Thermal stress due to non-radiative relaxation of the electron–hole pairs is the primary cause of the cooling. We also analyse...... an alternative cooling mechanism that is a result of electronic stress via the deformation potential, and outline future directions for cavity optomechanics with optically active semiconductors.......-quality-factor and optically active semiconductor nanomembrane. The cooling is a result of electron–hole generation by cavity photons. Consequently, the cooling factor depends on the optical wavelength, varies drastically in the vicinity of the semiconductor bandgap, and follows the excitonic absorption behaviour...

  4. Comparison between MAAP and ECART predictions of radionuclide transport throughout a French standard PWR reactor coolant system; Transport des radionucleides dans le circuit primaire d`un REP: comparaison des codes MAAP et ECART

    Energy Technology Data Exchange (ETDEWEB)

    Hervouet, C.; Ranval, W. [Electricite de France (EDF), 92 - Clamart (France); Parozzi, F.; Eusebi, M. [Ente Nazionale per l`Energia Elettrica, Rome (Italy)

    1996-04-01

    In the framework of a collaboration agreement between EDF and ENEL, the MAAP (Modular Accident Analysis Program) and ECART (ENEL Code for Analysis of radionuclide Transport) predictions about the fission product retention inside the reactor cooling system of a French PWR 1300 MW during a small Loss of Coolant Accident were compared. The volatile fission products CsI, CsOH, TeO{sub 2} and the structural materials, all of them released early by the core, are more retained in MAAP than in ECART. On the other hand, the non-volatile fission products, released later, are more retained in ECART than in MAAP, because MAAP does not take into account diffusion-phoresis: in fact, this deposition phenomenon is very significant when the molten core vaporizes the water of the vessel lower plenum. Centrifugal deposition in bends, that can be modeled only with ECART, slightly increases the whole retention in the circuit if it is accounted for. (authors). 18 refs., figs., tabs.

  5. Simulation Research on Decay Heat Removal System in Primary Loop of Pool-type Sodium-cooled Fast Reactor%池式钠冷快堆事故余热排出系统一回路仿真研究

    Institute of Scientific and Technical Information of China (English)

    姜博; 张智刚; 于洋; 陈广亮; 张志俭

    2015-01-01

    池式钠冷快堆事故余热排出系统采用了非能动工作原理,依靠液态钠及空气的自然对流排出堆芯余热。为研究事故工况下余热排出系统一回路的换热能力,基于 FORTRAN 语言,建立堆芯单通道及盒间流模型,采用全隐二阶迎风差分格式及改进的欧拉法离散求解,对事故余热排出系统一回路系统进行数值模拟,并对全厂断电事故进行仿真计算验证。结果表明:该程序能较好地反映事故余热排出系统瞬态变化过程,并可达到超实时仿真。%T he decay heat removal system in pool‐type sodium‐cooled fast reactor (PSFR) is the passive safety system ,which depends on the natural circulation of sodium and air to keep the reactor coolant cooled .In order to verify the characteristics of the heat transfer of decay heat removal system in primary loop for accident condition ,the core single‐channel model and the flow between fuel assemblies model were established to simulate the decay heat removal system of primary loop and testify the program on station blackout accident , by using fully‐implicit second‐order upwind scheme and ameliorative Eular method to solve the equations based on FORTRAN .The calculation results show that the program could reflect the transient characteristics of the decay heat removal system ,and it could reach excess real‐time simulation .

  6. Self pumping magnetic cooling

    Science.gov (United States)

    Chaudhary, V.; Wang, Z.; Ray, A.; Sridhar, I.; Ramanujan, R. V.

    2017-01-01

    Efficient thermal management and heat recovery devices are of high technological significance for innovative energy conservation solutions. We describe a study of a self-pumping magnetic cooling device, which does not require external energy input, employing Mn-Zn ferrite nanoparticles suspended in water. The device performance depends strongly on magnetic field strength, nanoparticle content in the fluid and heat load temperature. Cooling (ΔT) by ~20 °C and ~28 °C was achieved by the application of 0.3 T magnetic field when the initial temperature of the heat load was 64 °C and 87 °C, respectively. These experiments results were in good agreement with simulations performed with COMSOL Multiphysics. Our system is a self-regulating device; as the heat load increases, the magnetization of the ferrofluid decreases; leading to an increase in the fluid velocity and consequently, faster heat transfer from the heat source to the heat sink.

  7. Cooled particle accelerator target

    Science.gov (United States)

    Degtiarenko, Pavel V.

    2005-06-14

    A novel particle beam target comprising: a rotating target disc mounted on a retainer and thermally coupled to a first array of spaced-apart parallel plate fins that extend radially inwardly from the retainer and mesh without physical contact with a second array of spaced-apart parallel plate fins that extend radially outwardly from and are thermally coupled to a cooling mechanism capable of removing heat from said second array of spaced-apart fins and located within the first array of spaced-apart parallel fins. Radiant thermal exchange between the two arrays of parallel plate fins provides removal of heat from the rotating disc. A method of cooling the rotating target is also described.

  8. Water Cooled Mirror Design

    Energy Technology Data Exchange (ETDEWEB)

    Dale, Gregory E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Holloway, Michael Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Pulliam, Elias Noel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-03-30

    This design is intended to replace the current mirror setup being used for the NorthStar Moly 99 project in order to monitor the target coupon. The existing setup has limited movement for camera alignment and is difficult to align properly. This proposed conceptual design for a water cooled mirror will allow for greater thermal transfer between the mirror and the water block. It will also improve positioning of the mirror by using flexible vacuum hosing and a ball head joint capable of a wide range of motion. Incorporating this design into the target monitoring system will provide more efficient cooling of the mirror which will improve the amount of diffraction caused by the heating of the mirror. The process of aligning the mirror for accurate position will be greatly improved by increasing the range of motion by offering six degrees of freedom.

  9. Magnetic entropy and cooling

    DEFF Research Database (Denmark)

    Hansen, Britt Rosendahl; Kuhn, Luise Theil; Bahl, Christian Robert Haffenden

    2010-01-01

    Some manifestations of magnetism are well-known and utilized on an everyday basis, e.g. using a refrigerator magnet for hanging that important note on the refrigerator door. Others are, so far, more exotic, such as cooling by making use of the magnetocaloric eect. This eect can cause a change...... or nitrogen liquefaction or for room-temperature cooling. The magnetocaloric eect can further be used to determine phase transition boundaries, if a change in the magnetic state occurs at the boundary.In this talk, I will introduce the magnetocaloric eect (MCE) and the two equations, which characterize...... in the temperature of a magnetic material when a magnetic eld is applied or removed. For many years, experimentalists have made use of dilute paramagnetic materials to achieve milliKelvin temperatures by use of the magnetocaloric eect. Also, research is done on materials, which might be used for hydrogen, helium...

  10. Cooling Floor AC Systems

    Science.gov (United States)

    Jun, Lu; Hao, Ding; Hong, Zhang; Ce, Gao Dian

    The present HVAC equipments for the residential buildings in the Hot-summer-and-Cold-winter climate region are still at a high energy consuming level. So that the high efficiency HVAC system is an urgently need for achieving the preset government energy saving goal. With its advantage of highly sanitary, highly comfortable and uniform of temperature field, the hot-water resource floor radiation heating system has been widely accepted. This paper has put forward a new way in air-conditioning, which combines the fresh-air supply unit and such floor radiation system for the dehumidification and cooling in summer or heating in winter. By analyze its advantages and limitations, we found that this so called Cooling/ Heating Floor AC System can improve the IAQ of residential building while keep high efficiency quality. We also recommend a methodology for the HVAC system designing, which will ensure the reduction of energy cost of users.

  11. ANALISIS MODEL TERAS 3-DIMENSI UNTUK EVALUASI PARAMETER KRITIKALITAS REAKTOR PWR MAJU KELAS 1000 MW

    Directory of Open Access Journals (Sweden)

    Tagor Malem Sembiring

    2015-04-01

    Full Text Available Setelah kejadian Fukushima, penggunaan sistem keselamatan pasif menjadi persyaratan yang penting untuk PLTN. PLTN jenis PWR maju kelas 1000 yang didesain oleh Westinghouse, AP1000, memiliki fitur keselamatan pasif disamping sederhana dan modular. Sebelum memilih suatu PLTN, maka perlu dilakukan suatu evaluasi terhadap parameter desainnya. Salah satu parameter yang penting dalam keselamatan adalah kritikalitas teras. Permasalahan pokok dalam mengevaluasi parameter kritikalitas teras AP1000 tidak adanya data komposisi material SS304 dan H2O di daerah reflektor dan diameter penyerap SS304. Dengan demikian tujuan penelitian ini adalah mendapatkan model teras 3-dimensi AP1000 dan siap diaplikasikan dalam evaluasi parameter kritikalitas teras. Hasil perhitungan menunjukkan bahwa komposisi terbaik SS304 dan H2O di reflektor teras bagian atas dan bawah masing-masing 50 vol%, sedangkan diameter penyerap SS304 adalah 0,960 cm. Evaluasi konsentrasi boron kritis menunjukkan perbedaan yang signifikan dengan nilai desain. Meskipun penyebab utama dari perbedaan ini belum diketahui, akan tetapi dapat dibuktikan bahwa konsentrasi boron kritis sangat sensitif dengan densitas UO2. Untuk reaktivitas padam, reaktor AP1000 memiliki margin subkritikalitas teras yang besar untuk satu siklus operasi. Dengan demikian teras yang diusulkan dapat digunakan sebagai acuan untuk evaluasi parameter teras lainnya atau perangkat analitis lainnya dalam rangka mengevaluasi desain reaktor AP1000. Kata kunci: AP1000, kritikalitas, konsentrasi boron kritis, reaktivitas padam   After the Fukushima accident, the use of passive safety system becomes an important requirement for the nuclear power plant (NPP. The advanced PWR NPP with 1000 MW (electric class, designed by Westinghouse, AP1000, a reactor with the passive safety features as well as simple and modular. Before selecting a nuclear power plant, there should be an evaluation of the design parameter. One important parameter in

  12. Valve inlet fluid conditions for pressurizer safety and relief valves for B and W 177-FA and 205-FA plants. Final report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cartin, L.R.; Winks, R.W.; Merchent, J.W.; Brandt, R.T.

    1982-12-01

    The overpressurization transients for the Babcock and Wilcox Company's 177- and 205-FA units are reviewed to determine the range of fluid conditions expected at the inlet of pressurizer safety and relief valves. The final Safety Analysis Report, extended high-pressure injection, and cold overpressurization events are considered. The results of this review, presented in the form of tables and graphs, provide input to the PWR utilities in their justification that the fluid conditions under which their valve designs were tested as part of the EPRI PWR Safety and Relief Valve Test Program are representative of those expected in their unit(s).

  13. Heating, ventilation and cooling

    CSIR Research Space (South Africa)

    Osburn, L

    2009-02-01

    Full Text Available content and is evaporated by the air stream with less moisture. Enthalpy wheels are more effective at transferring energy between the air streams as both sensible and latent heat is transferred. Ground-Coupled Heat Exchanger Ground-coupled heat... with high diurnal temperature variations. Evaporative Coolers Evaporative coolers work on the concept that the evaporation of water has a cooling effect on its immediate environment due to the latent heat that it absorbs in order to evaporate...

  14. Cooled Ion Frequency Standard.

    Science.gov (United States)

    2014-09-26

    when the cooling laser is turned off, the ions are heated by: (1) background gas collisions and (2) a plasma heating process which may be " resonant ...causes heating in our Penning traps. One way resonant particle transport is mediated is by misalignm.nt between the trap’s magnetic and electric axis...using computer solutions. The trap of Fig. 1 is noteworthy because although the inner surfaces of the trap are machined with simple conical cuts, the

  15. A MATLAB-Linked Solver to Find Fuel Depletion in a PWR, a Suggested VVER-1000 Type

    Directory of Open Access Journals (Sweden)

    F. Faghihi

    2009-01-01

    Full Text Available Coupled first-order IVPs are frequently used in many parts of engineering and sciences. We present a “solver” including three computer programs which were joint with the MATLAB software to solve and plot solutions of the first-order coupled stiff or nonstiff IVPs. Some applications related to IVPs are given here using our MATLAB-linked solver. Muon catalyzed fusion in a D-T mixture is considered as a first dynamical example of the coupled IVPs. Then, we have focused on the fuel depletion in a suggested PWR including poisons burnups (xenon-135 and samarium-149, plutonium isotopes production, and uranium depletion.

  16. Approaches to analyze the bowing of German PWR fuel assemblies; Ansaetze zur Analyse des Biegeverhaltens deutscher DWR-Brennelemente

    Energy Technology Data Exchange (ETDEWEB)

    Boeke, H.; Bauer, R.; Bloemeling, F.; Lawall, R. [TUeV NORD SysTec GmbH und Co. KG, Hamburg (Germany)

    2012-11-01

    The analysis of the bowing behavior of PWR fuel elements is required in case of increased fuel element deformations that have been observed during the last years. In the contribution the following issues are discussed: fuel element properties (stiffness, constructive features), influence factors (guiding tubes, spacer), load transfer and its impact. Under consideration of external boundary conditions an evaluation scheme was developed, using analysis data (control rod drop time), friction force measurements, fuel element characteristics (fuel element deformation, bowing) and their ranking, and simulation models (fluid-structure interactions). The evaluation scheme allows the definition of appropriate measures. The suitability of the methodology was demonstrated.

  17. Derivation of correction factor to be applied for calculated results of PWR fuel isotopic composition by ORIGEN2 code

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya; Nomura, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Murazaki, Minoru [Tokyo Nuclear Service Inc., Tokyo (Japan); Mochizuki, Hiroki [The Japan Research Institute Ltd., Tokyo (Japan)

    2001-11-01

    For providing conservative PWR spent fuel compositions from the view point of nuclear criticality safety, correction factors applicable for result of burnup calculation by ORIGEN2 were evaluated. Its conservativeness was verified by criticality calculations using MVP. To calculate these correction factors, analyses of spent fuel isotopic composition data were performed by ORIGEN2. Maximum or minimum value of the ratio of calculation result to experimental data was chosen as correction factor. These factors are given to each set of fuel assembly and ORIGEN2 library. They could be considered as the re-definition of recommended isotopic composition given in Nuclear Criticality Safety Handbook. (author)

  18. On-line PWR RHR pump performance testing following motor and impeller replacement

    Energy Technology Data Exchange (ETDEWEB)

    DiMarzo, J.T.

    1996-12-01

    On-line maintenance and replacement of safety-related pumps requires the performance of an inservice test to determine and confirm the operational readiness of the pumps. In 1995, major maintenance was performed on two Pressurized Water Reactor (PWR) Residual Heat Removal (RHR) Pumps. A refurbished spare motor was overhauled with a new mechanical seal, new motor bearings and equipped with pump`s `B` impeller. The spare was installed into the `B` train. The motor had never been run in the system before. A pump performance test was developed to verify it`s operational readiness and determine the in-situ pump performance curve. Since the unit was operating, emphasis was placed on conducting a highly accurate pump performance test that would ensure that it satisfied the NSSS vendors accident analysis minimum acceptance curve. The design of the RHR System allowed testing of one train while the other was aligned for normal operation. A test flow path was established from the Refueling Water Storage Tank (RWST) through the pump (under test) and back to the RWST. This allowed staff to conduct a full flow range pump performance test. Each train was analyzed and an expression developed that included an error vector term for the TDH (ft), pressure (psig), and flow rate (gpm) using the variance error vector methodology. This method allowed the engineers to select a test instrumentation system that would yield accurate readings and minimal measurement errors, for data taken in the measurement of TDH (P,Q) versus Pump Flow Rate (Q). Test results for the `B` Train showed performance well in excess of the minimum required. The motor that was originally in the `B` train was similarly overhauled and equipped with `A` pump`s original impeller, re-installed in the `A` train, and tested. Analysis of the `A` train results indicate that the RHR pump`s performance was also well in excess of the vendors requirements.

  19. A safety and regulatory assessment of generic BWR and PWR permanently shutdown nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Travis, R.J.; Davis, R.E.; Grove, E.J.; Azarm, M.A. [Brookhaven National Lab., Upton, NY (United States)

    1997-08-01

    The long-term availability of less expensive power and the increasing plant modification and maintenance costs have caused some utilities to re-examine the economics of nuclear power. As a result, several utilities have opted to permanently shutdown their plants. Each licensee of these permanently shutdown (PSD) plants has submitted plant-specific exemption requests for those regulations that they believe are no longer applicable to their facility. This report presents a regulatory assessment for generic BWR and PWR plants that have permanently ceased operation in support of NRC rulemaking activities in this area. After the reactor vessel is defueled, the traditional accident sequences that dominate the operating plant risk are no longer applicable. The remaining source of public risk is associated with the accidents that involve the spent fuel. Previous studies have indicated that complete spent fuel pool drainage is an accident of potential concern. Certain combinations of spent fuel storage configurations and decay times, could cause freshly discharged fuel assemblies to self heat to a temperature where the self sustained oxidation of the zircaloy fuel cladding may cause cladding failure. This study has defined four spent fuel configurations which encompass all of the anticipated spent fuel characteristics and storage modes following permanent shutdown. A representative accident sequence was chosen for each configuration. Consequence analyses were performed using these sequences to estimate onsite and boundary doses, population doses and economic costs. A list of candidate regulations was identified from a screening of 10 CFR Parts 0 to 199. The continued applicability of each regulation was assessed within the context of each spent fuel storage configuration and the results of the consequence analyses.

  20. Reference neutron transport calculation note for Korea nuclear power plants with 3-loop PWR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byung Cheol; Chang, Ki Oak

    1997-05-01

    Reactor pressure vessel (RPV) steels are subjected to neutron irradiation at a temperature of about 290 deg C. This radiation exposure alters the mechanical properties, leading to a shift of the brittle-to-ductile transition temperature toward higher temperatures and to a diminution of the rupture energy as determined by Charpy V-notch tests. This radiation embrittlement is one of the important aging factors of nuclear power plants. U.S. NRC recommended the basic requirements for the determination of the pressure vessel fluence by regulatory guide DG-1025 in order to reduce the uncertainty in the determination of neutron fluence calculation and measurements. The determination of the pressure vessel fluence is based on both calculations and measurements. The fluence prediction is made with a calculation and the measurements are used to qualify the calculational methodology. Because of the importance and the difficulty of these calculations, the method`s qualification by comparison to measurement must be made to ensure a reliable and accurate vessel fluence determination. This reference calculation note is to provide a series of forward and adjoint neutron transport calculations for use in the evaluation of neutron dosimetry from surveillance capsule irradiations at 3-loop PWR reactor as well as for use in the determination of the neutron exposure of the reactor vessel wall in accordance with U.S Regulatory Guide DG-1025 requirements. The calculations of the pressure vessel fluence consist of the following steps; (1) Determination of the geometrical and material input data, (2) Determination of the core neutron source, and (3) Propagation of the neutron fluence from the core to the vessel and into the cavity. (author). 12 tabs., 3 figs., 7 refs.

  1. PWR core and spent fuel pool analysis using scale and nestle

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, J. E.; Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee, Knoxville, TN 37996-2300 (United States); St Clair, R.; Orr, D. [Duke Energy, 526 S. Church St, Charlotte, NC 28202 (United States)

    2012-07-01

    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

  2. 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.

  3. Towards a reference numerical scheme using MCNPX for PWR control rod tip fluence estimations

    Energy Technology Data Exchange (ETDEWEB)

    Ferroukhi, H.; Vasiliev, A. [Paul Scherrer Institut, CH-5232 Villigen-PSI (Switzerland); Dufresne, A. [Dept. of Physics, EPFL, 1015 Lausanne (Switzerland); Chawla, R. [Dept. of Physics, EPFL, 1015 Lausanne (Switzerland); Paul Scherrer Institut (Switzerland)

    2012-07-01

    Recent occurrences of cracks and fissures on the cladding tubes of PWR control rod (CR) fingers employed in the Swiss reactors prompted the need to develop more reliable analytical methods for CR tip fluence estimations. To partly address this need, a deterministic methodology based on SIMULATE-3/CASMO-4 was in recent years developed at PSI. Although this methodology has already been applied for independent support to licensing issues related to CR lifetime, two main questions are currently being the center of attention for further enhancements. First, the methodology relies on several assumptions that have so far not been verified. Secondly, an assessment of the achieved accuracy has not been addressed. In an attempt to answer both these open questions, it was considered appropriate to develop an alternative computational scheme based on the stochastic MCNPX code with the objective to provide reference numerical solutions. This paper presents the first steps undertaken in that direction. To start, a methodology for a volumetric neutron source transfer to full core MCNPX models with detailed CR as well as axial reflector representations is established. On this basis, the assumptions of the deterministic methodology are studied for selected CR configurations for two Beginning-of-Life cores by comparing the spatial neutron flux distributions obtained with the two approaches for the entire spectrum. Finally, for the high-energy range (E> 1 MeV) and for a few CRs, the new MCNPX scheme is applied to estimate the accumulated fluence over one real operated cycle and the results are compared with the deterministic approach. (authors)

  4. Nuclear Data Library Effects on Fast to Thermal Flux Shapes Around PWR Control Rod Tips

    Science.gov (United States)

    Vasiliev, A.; Ferroukhi, H.; Zhu, T.; Pautz, A.

    2014-04-01

    The development of a high-fidelity computational scheme to estimate the accumulated fluence at the tips of PWR control rods (CR) has been initiated at the Paul Scherrer Institut (PSI). Both the fluence from high-energy (E>1 MeV) neutrons as well as for the thermal range (E<0.625 eV) are required as these affect the CR integrity through stresses/strains induced by coupled clad embrittlement / absorber swelling phenomena. The concept of the PSI scheme under development is to provide from validated core analysis models, the volumetric neutron source to a full core MCNPX model that is then used to compute the neutron fluxes. A particular aspect that needs scrutiny is the ability of the MCNPX-based calculation methodology to accurately predict the flux shapes along the control rod surfaces, especially for fully withdrawn CRs. In that case, the tip is located a short distance above the core/reflector interface and since this situation corresponds to a large part of reactor operation, the accumulated fluence will highly depend on the achieved calculation accuracy and precision in this non-fueled zone. The objective of the work presented in this paper is to quantify the influence of nuclear data on the calculated fluxes at the CR tips by (1) conducting a systematic comparison of modern neutron cross-section libraries, including JENDL-4.0, JEFF-3.1.1 and ENDF/B-VII.0, and (2) by quantifying the uncertainties in the neutron flux calculations with the help of available neutron cross-section variances/covariances data. For completeness, the magnitude of these nuclear data-based uncertainties is also assessed in relation to the influence from other typical sources of modeling uncertainties/biases.

  5. Performance evaluation of PSO and GA in PWR core loading pattern optimization

    Energy Technology Data Exchange (ETDEWEB)

    Khoshahval, F., E-mail: f_khoshahval@sbu.ac.i [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Velenjak, Tehran (Iran, Islamic Republic of); Minuchehr, H. [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Velenjak, Tehran (Iran, Islamic Republic of); Zolfaghari, A., E-mail: a-zolfaghari@sbu.ac.i [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Velenjak, Tehran (Iran, Islamic Republic of)

    2011-03-15

    Research highlights: The performance of both GA and PSO methods in optimizing of a PWR core are adequate. It seems GA arrives to its final parameter value in a fewer generation than the PSO. The computation time for GA is higher than PSO. The GA-2 and PSO-CFA algorithms perform better in comparison to GA-1 and PSO-IWA. - Abstract: The efficient operation and fuel management of PWRs are of utmost importance. Recently, genetic algorithm (GA) and particle swarm optimization (PSO) techniques have attracted considerable attention among various modern heuristic optimization techniques. GA is a powerful optimization technique, based upon the principles of natural selection and species evolution. GA is finding popularity as design tools because of its versatility, intuitiveness and ability to solve highly non-linear, mixed integer optimization problems. PSO refers to a relatively new family of algorithms and is mainly inspired by social behavior patterns of organisms that live within large group. This study addresses the application and performance comparison of PSO and GA optimization methods for nuclear fuel loading pattern problem. Flattening of power inside the reactor core of Bushehr nuclear power plant (WWER-1000 type) is chosen as an objective function to prove the validity of algorithms. In addition the performance of both optimization techniques in terms of convergence rate and computational time is compared. It is found that, from an evolutionary point of view, the performance of both GA and PSO is quite adequate. But, GA seems to arrive at its final parameter value in a fewer generations than the PSO. It is also noticed that, the computation time for implemented GA in this work is too high in comparison to PSO.

  6. Applicability of 3D Monte Carlo simulations for local values calculations in a PWR core

    Science.gov (United States)

    Bernard, Franck; Cochet, Bertrand; Jinaphanh, Alexis; Jacquet, Olivier

    2014-06-01

    As technical support of the French Nuclear Safety Authority, IRSN has been developing the MORET Monte Carlo code for many years in the framework of criticality safety assessment and is now working to extend its application to reactor physics. For that purpose, beside the validation for criticality safety (more than 2000 benchmarks from the ICSBEP Handbook have been modeled and analyzed), a complementary validation phase for reactor physics has been started, with benchmarks from IRPHEP Handbook and others. In particular, to evaluate the applicability of MORET and other Monte Carlo codes for local flux or power density calculations in large power reactors, it has been decided to contribute to the "Monte Carlo Performance Benchmark" (hosted by OECD/NEA). The aim of this benchmark is to monitor, in forthcoming decades, the performance progress of detailed Monte Carlo full core calculations. More precisely, it measures their advancement towards achieving high statistical accuracy in reasonable computation time for local power at fuel pellet level. A full PWR reactor core is modeled to compute local power densities for more than 6 million fuel regions. This paper presents results obtained at IRSN for this benchmark with MORET and comparisons with MCNP. The number of fuel elements is so large that source convergence as well as statistical convergence issues could cause large errors in local tallies, especially in peripheral zones. Various sampling or tracking methods have been implemented in MORET, and their operational effects on such a complex case have been studied. Beyond convergence issues, to compute local values in so many fuel regions could cause prohibitive slowing down of neutron tracking. To avoid this, energy grid unification and tallies preparation before tracking have been implemented, tested and proved to be successful. In this particular case, IRSN obtained promising results with MORET compared to MCNP, in terms of local power densities, standard

  7. Severe accident modeling of a PWR core with different cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, S. C. [Westinghouse Electric Company LLC, 5801 Bluff Road, Columbia, SC 29209 (United States); Henry, R. E.; Paik, C. Y. [Fauske and Associates, Inc., 16W070 83rd Street, Burr Ridge, IL 60527 (United States)

    2012-07-01

    The MAAP v.4 software has been used to model two severe accident scenarios in nuclear power reactors with three different materials as fuel cladding. The TMI-2 severe accident was modeled with Zircaloy-2 and SiC as clad material and a SBO accident in a Zion-like, 4-loop, Westinghouse PWR was modeled with Zircaloy-2, SiC, and 304 stainless steel as clad material. TMI-2 modeling results indicate that lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would result if SiC was substituted for Zircaloy-2 as cladding. SBO modeling results indicate that the calculated time to RCS rupture would increase by approximately 20 minutes if SiC was substituted for Zircaloy-2. Additionally, when an extended SBO accident (RCS creep rupture failure disabled) was modeled, significantly lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would be generated by substituting SiC for Zircaloy-2 or stainless steel cladding. Because the rate of SiC oxidation reaction with elevated temperature H{sub 2}O (g) was set to 0 for this work, these results should be considered preliminary. However, the benefits of SiC as a more accident tolerant clad material have been shown and additional investigation of SiC as an LWR core material are warranted, specifically investigations of the oxidation kinetics of SiC in H{sub 2}O (g) over the range of temperatures and pressures relevant to severe accidents in LWR 's. (authors)

  8. Electron Cooling of RHIC

    CERN Document Server

    Ben-Zvi, Ilan; Barton, Donald; Beavis, Dana; Blaskiewicz, Michael; Bluem, Hans; Brennan, Joseph M; Bruhwiler, David L; Burger, Al; Burov, Alexey; Burrill, Andrew; Calaga, Rama; Cameron, Peter; Chang, Xiangyun; Cole, Michael; Connolly, Roger; Delayen, Jean R; Derbenev, Yaroslav S; Eidelman, Yury I; Favale, Anthony; Fedotov, Alexei V; Fischer, Wolfram; Funk, L W; Gassner, David M; Hahn, Harald; Harrison, Michael; Hershcovitch, Ady; Holmes, Douglas; Hseuh Hsiao Chaun; Johnson, Peter; Kayran, Dmitry; Kewisch, Jorg; Kneisel, Peter; Koop, Ivan; Lambiase, Robert; Litvinenko, Vladimir N; MacKay, William W; Mahler, George; Malitsky, Nikolay; McIntyre, Gary; Meng, Wuzheng; Merminga, Lia; Meshkov, Igor; Mirabella, Kerry; Montag, Christoph; Nagaitsev, Sergei; Nehring, Thomas; Nicoletti, Tony; Oerter, Brian; Parkhomchuk, Vasily; Parzen, George; Pate, David; Phillips, Larry; Preble, Joseph P; Rank, Jim; Rao, Triveni; Rathke, John; Roser, Thomas; Russo, Thomas; Scaduto, Joseph; Schultheiss, Tom; Sekutowicz, Jacek; Shatunov, Yuri; Sidorin, Anatoly O; Skrinsky, Aleksander Nikolayevich; Smirnov, Alexander V; Smith, Kevin T; Todd, Alan M M; Trbojevic, Dejan; Troubnikov, Grigory; Wang, Gang; Wei, Jie; Williams, Neville; Wu, Kuo-Chen; Yakimenko, Vitaly; Zaltsman, Alex; Zhao, Yongxiang; ain, Animesh K

    2005-01-01

    We report progress on the R&D program for electron-cooling of the Relativistic Heavy Ion Collider (RHIC). This electron cooler is designed to cool 100 GeV/nucleon at storage energy using 54 MeV electrons. The electron source will be a superconducting RF photocathode gun. The accelerator will be a superconducting energy recovery linac. The frequency of the accelerator is set at 703.75 MHz. The maximum electron bunch frequency is 9.38 MHz, with bunch charge of 20 nC. The R&D program has the following components: The photoinjector and its photocathode, the superconducting linac cavity, start-to-end beam dynamics with magnetized electrons, electron cooling calculations including benchmarking experiments and development of a large superconducting solenoid. The photoinjector and linac cavity are being incorporated into an energy recovery linac aimed at demonstrating ampere class current at about 20 MeV. A Zeroth Order Design Report is in an advanced draft state, and can be found on the web at http://www.ags...

  9. Low mass integrated cooling

    CERN Document Server

    Mapelli, Alessandro

    2014-01-01

    Low mass on - detec tor cooling systems are being developed and stud ied by the Detector Technology group (PH - DT) in the CERN Physics Department in close collaboration with LHC and non - LHC experiments . Two approaches are currently being investigated. The first approach, for barrel configurations, consists in integrating the cooli ng apparatus in light mechanical structures support ing the detectors. In this case , the thermal management can be achieved either with light cooling pipes and thin plates or with a network of microchannels embedded in thin strips of silicon or polyimide . Both configuratio ns are being investigated in the context of the 2018 upgrade program of the ALICE Inner Tracking System (ITS). Moreover, it is also possible to use a s ilicon microchannel cooling device itself as structural support for the detectors and electronics. Such a configur ation has been adopted by the NA62 collaboration for the ir GigaTracKer (GTK) as well as by the LHCb collaboration for the 2018 major upgrade of...

  10. Ultraefficient Cooling of Resonators: Beating Sideband Cooling with Quantum Control

    Science.gov (United States)

    Wang, Xiaoting; Vinjanampathy, Sai; Strauch, Frederick; Jacobs, Kurt

    2012-02-01

    There is presently a great deal of interest in cooling high-frequency micro- and nano-mechanical oscillators to their ground states. The present state of the art in cooling mechanical resonators is a version of sideband cooling, which was originally developed in the context of cooling trapped ions. Here we present a method based on quantum control that uses the same configuration as sideband cooling--coupling the resonator to be cooled to a second microwave (or optical) auxiliary resonator--but will cool significantly colder. This is achieved by applying optimal control and varying the strength of the coupling between the two resonators over a time on the order of the period of the mechanical resonator. As part of our analysis, we also obtain a method for fast, high-fidelity quantum information transfer between resonators.

  11. Thermal computations for electronics conductive, radiative, and convective air cooling

    CERN Document Server

    Ellison, Gordon

    2010-01-01

    IntroductionPrimary mechanisms of heat flowConductionApplication example: Silicon chip resistance calculationConvectionApplication example: Chassis panel cooled by natural convectionRadiationApplication example: Chassis panel cooled only by radiation 7Illustrative example: Simple thermal network model for a heat sinked power transistorIllustrative example: Thermal network circuit for a printed circuit boardCompact component modelsIllustrative example: Pressure and thermal circuits for a forced air cooled enclosureIllustrative example: A single chip package on a printed circuit board-the proble

  12. Cooling lubricants; Kuehlschmierstoffe

    Energy Technology Data Exchange (ETDEWEB)

    Pfeiffer, W. [Berufsgenossenschaftliches Inst. fuer Arbeitssicherheit, St. Augustin (Germany); Breuer, D. [Berufsgenossenschaftliches Inst. fuer Arbeitssicherheit, St. Augustin (Germany); Blome, H. [Berufsgenossenschaftliches Inst. fuer Arbeitssicherheit, St. Augustin (Germany); Deininger, C. [Berufsgenossenschaftliches Inst. fuer Arbeitssicherheit, St. Augustin (Germany); Hahn, J.U. [Berufsgenossenschaftliches Inst. fuer Arbeitssicherheit, St. Augustin (Germany); Kleine, H. [Berufsgenossenschaftliches Inst. fuer Arbeitssicherheit, St. Augustin (Germany); Nies, E. [Berufsgenossenschaftliches Inst. fuer Arbeitssicherheit, St. Augustin (Germany); Pflaumbaum, W. [Berufsgenossenschaftliches Inst. fuer Arbeitssicherheit, St. Augustin (Germany); Stockmann, R. [Berufsgenossenschaftliches Inst. fuer Arbeitssicherheit, St. Augustin (Germany); Willert, G. [Berufsgenossenschaftliches Inst. fuer Arbeitssicherheit, St. Augustin (Germany); Sonnenschein, G. [Maschinenbau- und Metall-Berufsgenossenschaft, Duesseldorf (Germany)

    1996-08-01

    As a rule, the base substances used are certain liquid hydrocarbons from mineral oils as well as from native and synthetic oils. Through the addition of further substances the cooling lubricant takes on the particular qualities required for the use in question. Employees working with cooling lubricants are exposed to various hazards. The assessment of the concentrations at the work station is carried out on the basis of existing technical rules for contact with hazardous substances. However, the application/implementation of compulsory investigation and supervision in accordance with these rules is made difficult by the fact that cooling lubricants are, as a rule, made up of complicated compound mixtures. In addition to protecting employees from exposure to mists and vapours from the cooling lubricants, protection for the skin is also of particular importance. Cooling lubricants should not, if at all possible, be brought into contact with the skin. Cleansing the skin and skin care is just as important as changing working clothes regularly, and hygiene and cleanliness at the workplace. Unavoidable emissions are to be immediately collected at the point where they arise or are released and safely disposed of. This means taking into account all sources of emissions. The programme presented in this report therefore gives a very detailed account of the individual protective measures and provides recommendations for the design of technical protection facilities. (orig./MG) [Deutsch] Als Basisstoffe dienen in der Regel bestimmte fluessige Kohlenwasserstoffverbindungen aus Mineraloelen sowie aus nativen oder synthetischen Oelen. Durch die Zugabe von weiteren Stoffen erlangt der Kuehlschmierstoff seine fuer den jeweiligen Anwendungsabfall geforderten Eigenschaften. Beschaeftigte, die mit Kuehlschmierstoffen umgehen, sind unterschiedliche Gefahren ausgesetzt. Die Beurteilung der Kuehlschmierstoffkonzentrationen in der Luft am Arbeitsplatz erfolgt auf der Grundlage bestehender

  13. Development of a feed-and-bleed operation strategy with hybrid-SIT under low pressure condition of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, In Seop, E-mail: jeoni@rpi.edu [Department of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, 110 8th Street, Troy, NY (United States); Han, Sang Hoon, E-mail: shhan2@kaeri.re.kr [Advanced Research Group, Korea Atomic Energy Research Institute, 70 Daedeok-daero 989 Beon-gil, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Kang, Sang Hee, E-mail: sanghee.kang@khnp.co.kr [NSSS Design Group, Korea Hydro & Nuclear Power Co., Ltd., Central Research Institute, 70, 1312-beongil, Yuseongdaero, Yuseong-gu, Daejeon (Korea, Republic of); Kang, Hyun Gook, E-mail: hyungook@kaist.ac.kr [Department of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, 110 8th Street, Troy, NY (United States)

    2017-04-01

    Highlights: • The novel F&B operation strategy with H-SIT and LPSI is developed. • The effectiveness of the H-SITs is verified using thermo-hydraulic simulations. • Success criteria considered for the new F&B operation strategy is identified. • A PSA model of APR+ reflecting the new F&B strategy with H-SIT is developed. • A risk analysis of the proposed F&B operation strategy is performed. - Abstract: While safety functions in current nuclear power plants are mainly provided by active safety systems, recently passive safety systems are being combined with the active systems to strengthen accident mitigation capability and therefore enhance overall plant safety. To this end, securing long-term cooling of the core is of particular importance. This study considers the hybrid safety injection tank (H-SIT), a passive injection system, as a target component to develop a long-term cooling strategy using active and passive systems concurrently. In the feed-and-bleed (F&B) operation, one of the important long-term cooling strategies to maintain core safety in pressurized water reactors, low pressure safety injection (LPSI) pumps are typically considered inoperable as depressurization is first required, which leads to core dry-out before reaching LPSI operable pressure. This study investigates whether H-SITs, with the important design feature of passive coolant injection under any pressure condition of the primary coolant system, can make up the core during depressurization thereby allowing LPSI pumps to be used in F&B operation as an additional means of long-term cooling. The effectiveness of the H-SITs is verified using thermal-hydraulic simulations, and based on the results a novel F&B operation strategy with H-SITs and LPSI pumps is developed. A probabilistic safety assessment (PSA) model is then developed in order to assess the risk effect of the suggested strategy. PSA results demonstrate that the proposed strategy lowers core damage frequency in the target

  14. Laser Cooling of Molecular Anions

    CERN Document Server

    Yzombard, Pauline; Gerber, Sebastian; Doser, Michael; Comparat, Daniel

    2015-01-01

    We propose a scheme for laser cooling of negatively charged molecules. We briefly summarise the requirements for such laser cooling and we identify a number of potential candidates. A detailed computation study with C$\\_2^-$, the most studied molecular anion, is carried out. Simulations of 3D laser cooling in a gas phase show that this molecule could be cooled down to below 1 mK in only a few tens of milliseconds, using standard lasers. Sisyphus cooling, where no photo-detachment process is present, as well as Doppler laser cooling of trapped C$\\_2^-$, are also simulated. This cooling scheme has an impact on the study of cold molecules, molecular anions, charged particle sources and antimatter physics.

  15. Laser cooling of molecular anions.

    Science.gov (United States)

    Yzombard, Pauline; Hamamda, Mehdi; Gerber, Sebastian; Doser, Michael; Comparat, Daniel

    2015-05-29

    We propose a scheme for laser cooling of negatively charged molecules. We briefly summarize the requirements for such laser cooling and we identify a number of potential candidates. A detailed computation study with C_{2}^{-}, the most studied molecular anion, is carried out. Simulations of 3D laser cooling in a gas phase show that this molecule could be cooled down to below 1 mK in only a few tens of milliseconds, using standard lasers. Sisyphus cooling, where no photodetachment process is present, as well as Doppler laser cooling of trapped C_{2}^{-}, are also simulated. This cooling scheme has an impact on the study of cold molecules, molecular anions, charged particle sources, and antimatter physics.

  16. Process integration: Cooling water systems design

    CSIR Research Space (South Africa)

    Gololo, KV

    2010-10-01

    Full Text Available This paper presents a technique for grassroot design of cooling water system for wastewater minimization which incorporates the performances of the cooling towers involved. The study focuses mainly on cooling systems consisting of multiple cooling...

  17. Cooling water systems design using process integration

    CSIR Research Space (South Africa)

    Gololo, KV

    2010-09-01

    Full Text Available Cooling water systems are generally designed with a set of heat exchangers arranged in parallel. This arrangement results in higher cooling water flowrate and low cooling water return temperature thus reducing cooling tower efficiency. Previous...

  18. Integrated functional modeling method for NPP plant DiD risk monitor and its application for conventional PWR

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Hidekazu; Yang, Ming; Zhang, Zhijian [Harbin Engineering University, Harbin (China)

    2014-08-15

    The development of a new risk monitor system is introduced in this paper, which can be applied not only to severe accident prevention in daily operation but also to serve as to mitigate the radiological hazard just after severe accident happens and long term management of post-severe accident consequences. The summary of the fundamental method is summarized on how to configure the Plant Defense in-Depth (Did) Risk Monitor by object-oriented software system based on functional modeling approach. Following the authors??preceding preliminary study for AP1000, the way of realizing the proposed method of configuring the plant Did risk monitor was investigated for a safety-enhanced Japanese PWR design to meet with the tight anti-severe accident requirements set by national regulation in Japan after Fukushima Daiichi accident. The result of this example practice of the presented preliminary study for Japanese PWR was for the level 4 of the Did in case of beyond design basis accident, that is, loss of all AC power + RCP seal LOCA, against the former case of AP1000 for level 3 Did in case of large LOCA.

  19. New long-cycle small modular PWR cores using particle type burnable poisons for low boron operation

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Hoseong; Hwang, Dae Hee [Department of Nuclear Engineering, Kyung Hee University, Deogyeong-daero, GiHeung-gu, Yongin, Gyeonggi-do 446-701 (Korea, Republic of); Hong, Ser Gi, E-mail: sergihong@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, Deogyeong-daero, GiHeung-gu, Yongin, Gyeonggi-do 446-701 (Korea, Republic of); Shin, Ho Choel [Core and Fuel Analysis Group, Korea Hydro & Nuclear Power Central Research Institute (KHNP-CRI), Daejon 305-343 (Korea, Republic of)

    2017-04-01

    Highlights: • New advanced burnable poison rods (BPR) are suggested for low boron operation in PWR. • The new SMR cores have long cycle length of ∼4.5 EFPYs with low boron concentration. • The SMR core satisfies all the design targets and constraints. - Abstract: In this paper, new small long-cycle PWR (Pressurized Water Reactor) cores for low boron concentration operation are designed by employing advanced burnable poison rods (BPRs) in which the BISO (Bi-Isotropic) particles of burnable poison are distributed in a SiC matrix. The BPRs are designed by adjusting the kernel diameter, the kernel material and the packing fraction to effectively reduce the excess reactivity in order to reduce the boron concentration in the coolant and achieve a flat change in excess reactivity over a long operational cycle. In addition, axial zoning of the BPRs was suggested to improve the core performances, and it was shown that the suggested axial zoning of BPRs considerably extends the cycle length compared to a core with no BPR axial zoning. The results of the core physics analyses showed that the cores using BPRs with a B{sub 4}C kernel have long cycle lengths of ∼4.5 EFPYs (Effective Full Power Years), small maximum CBCs (Critical Boron Concentration) lower than 370 ppm, low power peaking factors, and large shutdown margins of control element assemblies.

  20. Improvement of availability of PWR nuclear plants through the reduction of the time required for refueling/maintenance outages

    Energy Technology Data Exchange (ETDEWEB)

    Mayers, J.B.; Soth, L.G.

    1978-04-01

    The objective of the project, conducted by Commonwealth Research Corporation and Westinghouse Electric Corporation, is to identify improvements in procedures and equipment which will reduce the time required for refueling/maintenance outages at PWR nuclear power plants. The outage of Commonwealth Edison Zion Station Unit 1 in March through May of 1976 was evaluated to identify those items which caused delays and those work activities that offer the potential for significant improvements that could reduce the overall duration of the outage and achieve an improvement in the plant's availability for power production. Modifications in procedures have been developed and were evaluated during one or more outages in 1977. Conceptual designs have been developed for equipment modifications to the refueling system that could reduce the time required for the refueling portion of the outage. The purpose of the interim report is to describe those conceptual designs and to assess their impact upon future outages. Recommendations are included for the implementation of these equipment improvements in a continuation of this program as a demonstration of plant availability benefits that can be realized in PWR nuclear plants already in operation or under construction.

  1. Overview and Discussion of the OECD/NRC Benchmark Based on NUPEC PWR Subchannel and Bundle Tests

    Directory of Open Access Journals (Sweden)

    M. Avramova

    2013-01-01

    Full Text Available The Pennsylvania State University (PSU under the sponsorship of the US Nuclear Regulatory Commission (NRC has prepared, organized, conducted, and summarized the Organisation for Economic Co-operation and Development/US Nuclear Regulatory Commission (OECD/NRC benchmark based on the Nuclear Power Engineering Corporation (NUPEC pressurized water reactor (PWR subchannel and bundle tests (PSBTs. The international benchmark activities have been conducted in cooperation with the Nuclear Energy Agency (NEA of OECD and the Japan Nuclear Energy Safety Organization (JNES, Japan. The OECD/NRC PSBT benchmark was organized to provide a test bed for assessing the capabilities of various thermal-hydraulic subchannel, system, and computational fluid dynamics (CFDs codes. The benchmark was designed to systematically assess and compare the participants’ numerical models for prediction of detailed subchannel void distribution and department from nucleate boiling (DNB, under steady-state and transient conditions, to full-scale experimental data. This paper provides an overview of the objectives of the benchmark along with a definition of the benchmark phases and exercises. The NUPEC PWR PSBT facility and the specific methods used in the void distribution measurements are discussed followed by a summary of comparative analyses of submitted final results for the exercises of the two benchmark phases.

  2. Solar heating and cooling.

    Science.gov (United States)

    Duffie, J A; Beckman, W A

    1976-01-16

    We have adequate theory and engineering capability to design, install, and use equipment for solar space and water heating. Energy can be delivered at costs that are competitive now with such high-cost energy sources as much fuel-generated, electrical resistance heating. The technology of heating is being improved through collector developments, improved materials, and studies of new ways to carry out the heating processes. Solar cooling is still in the experimental stage. Relatively few experiments have yielded information on solar operation of absorption coolers, on use of night sky radiation in locations with clear skies, on the combination of a solar-operated Rankine engine and a compression cooler, and on open cycle, humidification-dehumidification systems. Many more possibilities for exploration exist. Solar cooling may benefit from collector developments that permit energy delivery at higher temperatures and thus solar operation of additional kinds of cycles. Improved solar cooling capability can open up new applications of solar energy, particularly for larger buildings, and can result in markets for retrofitting existing buildings. Solar energy for buildings can, in the next decade, make a significant contribution to the national energy economy and to the pocketbooks of many individual users. very large-aggregate enterprises in manufacture, sale, and installation of solar energy equipment can result, which can involve a spectrum of large and small businesses. In our view, the technology is here or will soon be at hand; thus the basic decisions as to whether the United States uses this resource will be political in nature.

  3. Laser Cooling of Solids

    Science.gov (United States)

    2009-01-01

    observed in a range of glasses and crystals doped with Yb3+ (ZBLANP [19–22], ZBLAN [23,24], CNBZn [9,25] BIG [25, 26], KGd(WO4)2 [9], KY(WO4)2 [9], YAG [27...Yb3+-doped fluorozirconate glass ZBLAN , Phys. Rev. B 75, 144302 (2007). [40] C. W. Hoyt, Laser Cooling in Thulium-doped Solids, Ph. D. Thesis...date, optical refrigeration research has been confined to glasses and crystals doped with rare- earth elements and direct-band semiconductors such as

  4. Methodology to monitor and diagnostic vibrations of the motor-pumps used in the primary cooling system of IEAR-1 nuclear research reactor; Metodologia para monitoracao e diagnostico de vibracao das bombas moto-operadas do circuito primario de refrigeracao do Reator IEA-R1

    Energy Technology Data Exchange (ETDEWEB)

    Benevenuti, Erion de Lima

    2004-07-01

    The objectives of this study are to establish a strategy to monitor and diagnose vibrations of the motor pumps used in the primary reactor cooling system of the IEA-R1 nuclear research reactor, to verify the possibility of using the existing installed monitoring vibration system and to implement such strategy in a continuous way. Four types of mechanical problems were considered: unbalancing, misalignment, gaps and faults in bearings. An adequate set of analysis tools, well established by the industry, was selected. These are: global measurements of vibration, velocity spectrum and acceleration envelope spectrum. Three sources of data and information were used; the data measured from the primary pumps, experimental results obtained with a Spectra Quest machine used to simulate mechanical defects and data from the literature. The results show that, for the specific case of the motor-pumps of IEA-R1 nuclear research reactor, although the technique using the envelope of acceleration, which is not available in the current system used to monitor the vibration of the motor pumps, is the one with best performance, the other techniques available in the system are sufficient to monitor the four types of mechanical problems mentioned. The proposed strategy is shown and detailed in this work. (author)

  5. Role of hydrogen in the intergranular cracking mechanism by stress corrosion in primary medium of nickel based alloys 600 and 690; Role de l'hydrogene dans le mecanisme de fissuration intergranulaire par corrosion sous contrainte en milieu primaire des alliages base nickel 600, 690

    Energy Technology Data Exchange (ETDEWEB)

    Odemer, G.; Coudurier, A.; Jambon, F.; Chene, J. [CEA Saclay, Dept. de Physico-Chimie (DEN/DANS/DPC/SCCME/LECA), 91 - Gif sur Yvette (France); Odemer, G.; Coudurier, A.; Chene, J. [Evry Univ., UMR 8587 CNRS / CEA, LAMBE, 91 (France)

    2007-07-01

    The aim of this work is to characterize the sensitivity to hydrogen embrittlement of alloys 600 and 690 in order to better understand the eventual role of hydrogen in the stress corrosion mechanism which affects these alloys when they are exposed in PWR primary medium. (O.M.)

  6. RAMI analysis for DEMO HCPB blanket concept cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Dongiovanni, Danilo N., E-mail: danilo.dongiovanni@enea.it [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati (Italy); Pinna, Tonio [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati (Italy); Carloni, Dario [KIT, Institute of Neutron Physics and Reactor Technology (INR) – KIT (Germany)

    2015-10-15

    Highlights: • RAMI (reliability, availability, maintainability and inspectability) preliminary assessment for HCPB blanket concept cooling system. • Reliability block diagram (RBD) modeling and analysis for HCPB primary heat transfer system (PHTS), coolant purification system (CPS), pressure control system (PCS), and secondary cooling system. • Sensitivity analysis on system availability performance. • Failure models and repair models estimated on the base of data from the ENEA fusion component failure rate database (FCFRDB). - Abstract: A preliminary RAMI (reliability, availability, maintainability and inspectability) assessment for the HCPB (helium cooled pebble bed) blanket cooling system based on currently available design for DEMO fusion power plant is presented. The following sub-systems were considered in the analysis: blanket modules, primary cooling loop including pipework and steam generators lines, pressure control system (PCS), coolant purification system (CPS) and secondary cooling system. For PCS and CPS systems an extrapolation from ITER Test Blanket Module corresponding systems was used as reference design in the analysis. Helium cooled pebble bed (HCPB) system reliability block diagrams (RBD) models were implemented taking into account: system reliability-wise configuration, operating schedule currently foreseen for DEMO, maintenance schedule and plant evolution schedule as well as failure and corrective maintenance models. A simulation of plant activity was then performed on implemented RBDs to estimate plant availability performance on a mission time of 30 calendar years. The resulting availability performance was finally compared to availability goals previously proposed for DEMO plant by a panel of experts. The study suggests that inherent availability goals proposed for DEMO PHTS system and Tokamak auxiliaries are potentially achievable for the primary loop of the HCPB concept cooling system, but not for the secondary loop. A

  7. PERFORMANCE ANALYSIS OF MECHANICAL DRAFT COOLING TOWER

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S; Alfred Garrett, A; James02 Bollinger, J; Larry Koffman, L

    2009-02-10

    Industrial processes use mechanical draft cooling towers (MDCT's) to dissipate waste heat by transferring heat from water to air via evaporative cooling, which causes air humidification. The Savannah River Site (SRS) has cross-flow and counter-current MDCT's consisting of four independent compartments called cells. Each cell has its own fan to help maximize heat transfer between ambient air and circulated water. The primary objective of the work is to simulate the cooling tower performance for the counter-current cooling tower and to conduct a parametric study under different fan speeds and ambient air conditions. The Savannah River National Laboratory (SRNL) developed a computational fluid dynamics (CFD) model and performed the benchmarking analysis against the integral measurement results to accomplish the objective. The model uses three-dimensional steady-state momentum, continuity equations, air-vapor species balance equation, and two-equation turbulence as the basic governing equations. It was assumed that vapor phase is always transported by the continuous air phase with no slip velocity. In this case, water droplet component was considered as discrete phase for the interfacial heat and mass transfer via Lagrangian approach. Thus, the air-vapor mixture model with discrete water droplet phase is used for the analysis. A series of parametric calculations was performed to investigate the impact of wind speeds and ambient conditions on the thermal performance of the cooling tower when fans were operating and when they were turned off. The model was also benchmarked against the literature data and the SRS integral test results for key parameters such as air temperature and humidity at the tower exit and water temperature for given ambient conditions. Detailed results will be published here.

  8. An analytical model for the prediction of fluid-elastic forces in a rod bundle subjected to axial flow: theory, experimental validation and application to PWR fuel assemblies; Calcul des forces fluidelastiques dans les faisceaux de tubes sous ecoulement axial: theorie, validation, application aux assemblages combustibles des REP

    Energy Technology Data Exchange (ETDEWEB)

    Beaud, F. [Electricite de France (EDF), 78 - Chatou (France)

    1997-12-31

    A model predicting the fluid-elastic forces in a bundle of circular cylinders subjected to axial flow is presented in this paper. Whereas previously published models were limited to circular flow channel, the present one allows to take a rectangular flow external boundary into account. For that purpose, an original approach is derived from the standard method of images. This model will eventually be used to predict the fluid-structure coupling between the flow of primary coolant and a fuel assemblies in PWR nuclear reactors. It is indeed of major importance since the flow is shown to induce quite high damping and could therefore mitigate the incidence of an external load like a seismic excitation on the dynamics of the assemblies. The proposed model is validated on two cases from the literature but still needs further comparisons with the experiments being currently carried out on the EDF set-up. The flow has been shown to induce an approximate 12% damping on a PWR fuel assembly, at nominal reactor conditions. The possible grid effect on the fluid-structure coupling has been neglected so far but will soon be investigated at EDF. (author). 16 refs.

  9. Example Calculations of In{sub v}essel Steam Explosions for a Prototypical PWR

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ik Kyu; Hong, Seong Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    In this paper, the sample calculation for the in{sub v}essel steam explosions were done by using the MC3D code. The evaluation of the computational code had been done against TROI experiments and the code had been adapted to a PWR ex{sub v}essel steam explosion calculations. MC3D is a code for the calculation of different types of multiphase multi-component flows. It has been built with the fuel-coolant interaction calculations in mind. It is, however, able to calculate very different situations and has a rather wide field of potential applications. MC3D is a set of two fuel-coolant interaction codes with a common numeric solver, one for the premixing phase and one for the explosion phase. In general, the steam explosion simulation with MC3D is being carried out in two steps. In the first step, the distributions of the melt, water, and vapor phases at steam explosion triggering are being calculated with the premixing module. These premixing simulation results present the input for the second step when the escalation and propagation of the steam explosion through the premixture are being calculated with the explosion module. The MC3D premixing model is a six-field application in which the melt is described by three fields. The first one is called 'continuous' and can describe many situations as, e.g., a jet or the melt lying on the bottom of a vessel. The second field corresponds to the droplets issued from the jet fragmentation. This field describes the discontinuous state of the fuel. The third field is optional and describes the fuel fragments issuing from drop fine fragmentation. The remaining three fields are the water, the vapor, and a noncondensable gas. The drop surface area is modeled with a standard interfacial area transport equation. In the explosion model, the continuous phase is not present and only the two fields related to the dispersed fuel are considered

  10. Preliminary assessment of a combined passive safety system for typical 3-loop PWR CPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Zijiang; Shan, Jianqiang, E-mail: jqshan@mail.xjtu.edu.cn; Gou, Junli

    2017-03-15

    Highlights: • A combined passive safety system was placed on a typical 3-loop PWR CPR1000. • Three accident analyses show the three different accident mitigation methods of the passive safety system. • The three mitigation methods were proved to be useful. - Abstract: As the development of the nuclear industry, passive technology turns out to be a remarkable characteristic of advanced nuclear power plants. Since the 20th century, much effort has been given to the passive technology, and a number of evolutionary passive systems have developed. Thoughts have been given to upgrade the existing reactors with passive systems to meet stricter safety demands. In this paper, the CPR1000 plant, which is one kind of mature pressurized water reactor plants in China, is improved with some passive systems to enhance safety. The passive systems selected are as follows: (1) the reactor makeup tank (RMT); (2) the advanced accumulator (A-ACC); (3) the in-containment refueling water storage tank (IRWST); (4) the passive emergency feed water system (PEFS), which is installed on the secondary side of SGs; (5) the passive depressurization system (PDS). Although these passive components is based on the passive technology of some advanced reactors, their structural and trip designs are adjusted specifically so that it could be able to mitigate accidents of the CPR1000. Utilizing the RELAP5/MOD3.3 code, accident analyses (small break loss of coolant accident, large break loss of coolant accident, main feed water line break accident) of this improved CPR1000 plant were presented to demonstrate three different accident mitigation methods of the safety system and to test whether the passive safety system preformed its function well. In the SBLOCA, all components of the passive safety system were put into work sequentially, which prevented the core uncover. The LBLOCA analysis illustrates the contribution of the A-ACCs whose small-flow-rate injection can control the maximum cladding

  11. Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1998-09-01

    The objective of this topical report is to present to the NRC for review and acceptance a methodology for using burnup credit in the design of criticality control systems for PWR spent fuel transportation packages, while maintaining the criticality safety margins and related requirements of 10 CFR Part 71 and 72. The proposed methodology consists of five major steps as summarized below: (1) Validate a computer code system to calculate isotopic concentrations in SNF created during burnup in the reactor core and subsequent decay. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). and (5) Verify that SNF assemblies meet the package loading criteria and confirm proper fuel assembly selection prior to loading. (This step is required but the details are outside the scope of this topical report.) When reviewed and accepted by the NRC, this topical report will serve as a criterion document for criticality control analysts and will provide steps for the use of actinide-only burnup credit in the design of criticality control systems. The NRC-accepted burnup credit methodology will be used by commercial SNF storage and transportation package designers. Design-specific burnup credit criticality analyses will be defined, developed, and documented in the Safety Analysis Report (SAR) for each specific storage or transportation package that uses burnup credit. These SARs will then be submitted to the NRC for review and approval. This topical report is expected to be referenced in a number of storage and transportation cask applications to be submitted by commercial cask and canister designers to the NRC. Therefore, NRC acceptance of this topical report will result in increased efficiency of the

  12. THE COOLING OF CORONAL PLASMAS. IV. CATASTROPHIC COOLING OF LOOPS

    Energy Technology Data Exchange (ETDEWEB)

    Cargill, P. J. [Space and Atmospheric Physics, The Blackett Laboratory, Imperial College, London SW7 2BW (United Kingdom); Bradshaw, S. J., E-mail: p.cargill@imperial.ac.uk [Department of Physics and Astronomy, Rice University, Houston, TX 77005 (United States)

    2013-07-20

    We examine the radiative cooling of coronal loops and demonstrate that the recently identified catastrophic cooling is due to the inability of a loop to sustain radiative/enthalpy cooling below a critical temperature, which can be >1 MK in flares, 0.5-1 MK in active regions, and 0.1 MK in long tenuous loops. Catastrophic cooling is characterized by a rapid fall in coronal temperature, while the coronal density changes by a small amount. Analytic expressions for the critical temperature are derived and show good agreement with numerical results. This effect considerably limits the lifetime of coronal plasmas below the critical temperature.

  13. Cooling Performance of an Impingement Cooling Device Combined with Pins

    Institute of Scientific and Technical Information of China (English)

    Dongliang QUAN; Songling LIU; Jianghai LI; Gaowen LIU

    2005-01-01

    Experimental study and one dimensional model analysis were conducted to investigate cooling performance of an integrated impingement and pin fin cooling device. A typical configuration specimen was made and tested in a large scale low speed closed-looped wind tunnel. Detailed two-dimensional contour maps of the temperature and cooling effectiveness were obtained for different pressure ratios and therefore different coolant flow-rates through the tested specimen. The experimental results showed that very high cooling effectiveness can be achieved by this cooling device with relatively small amount of coolant flow. Based on the theory of transpiration cooling in porous material, a one dimensional heat transfer model was established to analyze the effect of various parameters on cooling effectiveness. It was found from this model that the variation of heat transfer on the gas side, including heat transfer coefficient and film cooling effectiveness, of the specimen created much more effect on its cooling effectiveness than that of the coolant side. The predictions of the one-dimensional mode were compared and agreed well with the experimental data.

  14. How the diffuse Universe cools

    CERN Document Server

    Bertone, Serena; Schaye, Joop

    2013-01-01

    In this work we investigate the cooling channels of diffuse gas (i.e. n_H<0.1 cm^-3) in cosmology. We aim to identify the wavelengths where most of the energy is radiated in the form of emission lines or continuum radiation, and the main elements and ions responsible for the emission. We use a subset of cosmological, hydrodynamical runs from the OWLS project to calculate the emission of diffuse gas and its evolution with time. We find that at z=0 (z=2) about 70 (80) per cent of the energy emitted by diffuse gas is carried by emission lines, with the continuum radiation contributing the remainder. Hydrogen lines in the Lyman series are the primary contributors to the line emission, with a share of 16 (20) per cent. Oxygen lines are the main metal contributors at high redshift, while silicon, carbon and iron lines are strongest at low redshift, when the contributions of AGB stars and supernova Ia explosions to the metal budget become important and when there is more hot gas. The ionic species carrying the mo...

  15. Electronic cooling using thermoelectric devices

    Energy Technology Data Exchange (ETDEWEB)

    Zebarjadi, M., E-mail: m.zebarjadi@rutgers.edu [Department of Mechanical and Aerospace Engineering, Rutgers University, Piscataway, New Jersey 08854 (United States); Institute of Advanced Materials, Devices, and Nanotechnology, Rutgers University, Piscataway, New Jersey 08854 (United States)

    2015-05-18

    Thermoelectric coolers or Peltier coolers are used to pump heat in the opposite direction of the natural heat flux. These coolers have also been proposed for electronic cooling, wherein the aim is to pump heat in the natural heat flux direction and from hot spots to the colder ambient temperature. In this manuscript, we show that for such applications, one needs to use thermoelectric materials with large thermal conductivity and large power factor, instead of the traditionally used high ZT thermoelectric materials. We further show that with the known thermoelectric materials, the active cooling cannot compete with passive cooling, and one needs to explore a new set of materials to provide a cooling solution better than a regular copper heat sink. We propose a set of materials and directions for exploring possible materials candidates suitable for electronic cooling. Finally, to achieve maximum cooling, we propose to use thermoelectric elements as fins attached to copper blocks.

  16. Optical stochastic cooling in Tevatron

    CERN Document Server

    Lebedev, V

    2012-01-01

    Intrabeam scattering is the major mechanism resulting in a growth of beam emittances and fast luminosity degradation in the Tevatron. As a result in the case of optimal collider operation only about 40% of antiprotons are used to the store end and the rest are discarded. Beam cooling is the only effective remedy to increase the particle burn rate and, consequently, the luminosity. Unfortunately neither electron nor stochastic cooling can be effective at the Tevatron energy and bunch density. Thus the optical stochastic cooling (OSC) is the only promising technology capable to cool the Tevatron beam. Possible ways of such cooling implementation in the Tevatron and advances in the OSC cooling theory are discussed in this paper. The technique looks promising and potentially can double the average Tevatron luminosity without increasing its peak value and the antiproton production.

  17. Electron Cooling Experiments in CSR

    CERN Document Server

    Xiaodong, Yang

    2011-01-01

    The six species heavy ion beam was accumulated with the help of electron cooling in the main ring of Cooler Storage Ring of Heavy Ion Research Facility in Lanzhou(HIRFL-CSR), the ion beam accumulation dependence on the parameters of cooler was investigated experimentally. The 400MeV/u 12C6+ and 200MeV/u 129Xe54+ was stored and cooled in the experimental ring CSRe, the cooling force was measured in different condition.

  18. Radiative cooling for thermophotovoltaic systems

    Science.gov (United States)

    Zhou, Zhiguang; Sun, Xingshu; Bermel, Peter

    2016-09-01

    Radiative cooling has recently garnered a great deal of attention for its potential as an alternative method for photovoltaic thermal management. Here, we will consider the limits of radiative cooling for thermal management of electronics broadly, as well as a specific application to thermal power generation. We show that radiative cooling power can increase rapidly with temperature, and is particularly beneficial in systems lacking standard convective cooling. This finding indicates that systems previously operating at elevated temperatures (e.g., 80°C) can be passively cooled close to ambient under appropriate conditions with a reasonable cooling area. To examine these general principles for a previously unexplored application, we consider the problem of thermophotovoltaic (TPV) conversion of heat to electricity via thermal radiation illuminating a photovoltaic diode. Since TPV systems generally operate in vacuum, convective cooling is sharply limited, but radiative cooling can be implemented with proper choice of materials and structures. In this work, realistic simulations of system performance are performed using the rigorous coupled wave analysis (RCWA) techniques to capture thermal emitter radiation, PV diode absorption, and radiative cooling. We subsequently optimize the structural geometry within realistic design constraints to find the best configurations to minimize operating temperature. It is found that low-iron soda-lime glass can potentially cool the PV diode by a substantial amount, even to below ambient temperatures. The cooling effect can be further improved by adding 2D-periodic photonic crystal structures. We find that the improvement of efficiency can be as much as an 18% relative increase, relative to the non-radiatively cooled baseline, as well as a potentially significant improvement in PV diode lifetime.

  19. To Be Cool or Uncool?

    Institute of Scientific and Technical Information of China (English)

    袁会珍

    2007-01-01

    The western world has always been divided into two types of people-the cool and the uncool. It is a division that __1__ in school. The cool kids are good at __2__. They are __3__ with the opposite sex. They are good-looking and people want to __4__ their style. They can do their homework but they don't make a big effort. That would __5__ be cool.

  20. Electron cooling experiments in CSR

    Institute of Scientific and Technical Information of China (English)

    PARKHOMCHUK; Vasily; REVA; Vladimir

    2011-01-01

    The six species heavy ion beam was accumulated with the help of electron cooling in the main ring of Cooler Storage Ring of Heavy Ion Research Facility in Lanzhou (HIRFL-CSR). The ion beam accumulation dependence on the parameters of cooler was investigated experimentally. The 400 MeV/u 12C6+ and 200 MeV/u 129Xe54+ were stored and cooled in the experimental ring CSRe, and the cooling force was measured in different conditions.