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Sample records for typical pwr primary

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

    International Nuclear Information System (INIS)

    Tanaka, T.; Shimizu, S.; Ogata, Y.

    1997-01-01

    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

  2. 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.

  3. 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.

  4. 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.

  5. 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.

    2017-04-15

    This paper presents the radiation shielding model of a typical PWR (CNPP-II) at Chashma, Pakistan. The model was developed using Monte Carlo N Particle code [2], equipped with ENDF/B-VI continuous energy cross section libraries. This model was applied to calculate the neutron and gamma flux and dose rates in the radial direction at core mid plane. The simulated results were compared with the reference results of Shanghai Nuclear Engineering Research and Design Institute (SNERDI).

  6. Simplified model of a PWR primary circuit

    International Nuclear Information System (INIS)

    Souza, A.L.; Faya, A.J.G.

    1988-07-01

    The computer program RENUR was developed to perform a very simplified simulation of a typical PWR primary circuit. The program has mathematical models for the thermal-hydraulics of the reactor core and the pressurizer, the rest of the circuit being treated as a single volume. Heat conduction in the fuel rod is analyzed by a nodal model. Average and hot channels are treated so that bulk response of the core and DNBR can be evaluated. A homogenenous model is employed in the pressurizer. Results are presented for a steady-state situation as well as for a loss of load transient. Agreement with the results of more elaborate computer codes is good with substantial reduction in computer costs. (author) [pt

  7. EPRI PWR primary water chemistry guidelines revision

    International Nuclear Information System (INIS)

    McElrath, Joel; Fruzzetti, Keith

    2014-01-01

    EPRI periodically updates the PWR Primary Water Chemistry Guidelines as new information becomes available and as required by NEI 97-06 (Steam Generator Program Guidelines) and NEI 03-08 (Guideline for the Management of Materials Issues). The last revision of the PWR water chemistry guidelines identified an optimum primary water chemistry program based on then-current understanding of research and field information. This new revision provides further details with regard to primary water stress corrosion cracking (PWSCC), fuel integrity, and shutdown dose rates. A committee of industry experts, including utility specialists, nuclear steam supply system (NSSS) and fuel vendor representatives, Institute of Nuclear Power Operations (INPO) representatives, consultants, and EPRI staff collaborated in reviewing the available data on primary water chemistry, reactor water coolant system materials issues, fuel integrity and performance issues, and radiation dose rate issues. From the data, the committee updated the water chemistry guidelines that all PWR nuclear plants should adopt. The committee revised guidance with regard to optimization to reflect industry experience gained since the publication of Revision 6. Among the changes, the technical information regarding the impact of zinc injection on PWSCC initiation and dose rate reduction has been updated to reflect the current level of knowledge within the industry. Similarly, industry experience with elevated lithium concentrations with regard to fuel performance and radiation dose rates has been updated to reflect data collected to date. Recognizing that each nuclear plant owner has a unique set of design, operating, and corporate concerns, the guidelines committee has retained a method for plant-specific optimization. Revision 7 of the Pressurized Water Reactor Primary Water Chemistry Guidelines provides guidance for PWR primary systems of all manufacture and design. The guidelines continue to emphasize plant

  8. SARDAN- A program for the transients simulation in a typical PWR plant

    International Nuclear Information System (INIS)

    Mattos Santos, R.L.P. de.

    1979-10-01

    A program in FORTRAN-IV language was developed that simulates the behaviour of the primary circuit in a typical PWR plant during condition II transients, in particular uncontrolled withdrawal of a control rod set, control rod set drops and uncontrolled boron dilution. It the mathematical model adopted the reactor core, the hot piping to which a pressurizer is coupled, the steam generator and the cold piping are considered. The results obtained in the analysis of the mentioned accidents are compared to those present at the Final Safety Analysis Report (FSAR) of the Angra-1 reactor and are considered satisfactory. (F.E.) [pt

  9. Simplified model of a PWR primary coolant circuit

    International Nuclear Information System (INIS)

    Souza, A.L. de; Faya, A.J.G.

    1988-01-01

    The computer program RENUR was developed to perform a very simplified simulation of a typical PWR primary circuit. The program has mathematical models for the thermal-hydraulics of the reactor core and the pressurizer, the rest of the circuit being treated as a single volume. Heat conduction in the fuel rod is analysed by a nodal model. Average and hot channels are treated so that the bulk response of the core and DNBR can be evaluated. A Homogenenous model is employed in the pressurizer. Results are presented for a steady-state situation as well as for a loss of load transient. Agreement with the results of more elaborate computer codes is good with substantial reduction in computer costs. (author) [pt

  10. Thermohydraulic calculations of PWR primary circuits

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1984-01-01

    Some mathematical and numerical models from Retran computer codes aiming to simulate reactor transients, are presented. The equations used for calculating one-dimensional flow are integrated using mathematical methods from Flash code, with steam code to correlate the variables from thermodynamic state. The algorithm obtained was used for calculating a PWR reactor. (E.G.) [pt

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

    Energy Technology Data Exchange (ETDEWEB)

    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.).

  12. Data assimilation and PWR primary measurement

    International Nuclear Information System (INIS)

    Mercier, Thibaud

    2015-01-01

    A Pressurized Water Reactor (PWR) Reactor Coolant System (RCS) is a highly complex physical process: heterogeneous power, flow and temperature distributions are difficult to be accurately measured, since instrumentations are limited in number, thus leading to the relevant safety and protection margins. EDF R and D is seeking to assess the potential benefits of applying Data Assimilation to a PWR's RCS (Reactor Coolant System) measurements, in order to improve the estimators for parameters of a reactor's operating setpoint, i.e. improving accuracy and reducing uncertainties and biases of measured RCS parameters. In this thesis, we define a 0D semi-empirical model for RCS, satisfying the description level usually chosen by plant operators, and construct a Monte-Carlo Method (inspired from Ensemble Methods) in order to use this model with Data Assimilation tools. We apply this method on simulated data in order to assess the reduction of uncertainties on key parameters: results are beyond expectations, however strong hypothesis are required, implying a careful preprocessing of input data. (author)

  13. Substitution of cobalt alloying in PWR primary circuit gate valves

    International Nuclear Information System (INIS)

    Cachon, L.; Sudreau, F.; Brunel, L.

    1995-01-01

    The object of this study is qualify cobalt-free alternative alloys for valve applications. This paper focus on tribological characterization of numerous coatings is done by using the first one, of a classical type. Then tests are performed with the second one which simulates solicitations supported by gate valves in primary circuit of PWR. 35% Ni-Cr - 65% Cr 3 C 2 coating, deposited by detonation gun technology, gives us hope to find a substitute of Stelite 6. (author). 5 refs., 16 figs., 2 tabs

  14. Comparative analysis of station blackout accident progression in typical PWR, BWR, and PHWR

    International Nuclear Information System (INIS)

    Park, Soo Young; Ahn, Kwang Il

    2012-01-01

    Since the crisis at the Fukushima plants, severe accident progression during a station blackout accident in nuclear power plants is recognized as a very important area for accident management and emergency planning. The purpose of this study is to investigate the comparative characteristics of anticipated severe accident progression among the three typical types of nuclear reactors. A station blackout scenario, where all off-site power is lost and the diesel generators fail, is simulated as an initiating event of a severe accident sequence. In this study a comparative analysis was performed for typical pressurized water reactor (PWR), boiling water reactor (BWR), and pressurized heavy water reactor (PHWR). The study includes the summarization of design differences that would impact severe accident progressions, thermal hydraulic/severe accident phenomenological analysis during a station blackout initiated-severe accident; and an investigation of the core damage process, both within the reactor vessel before it fails and in the containment afterwards, and the resultant impact on the containment.

  15. SCC of Alloy 600 components in PWR primary loop

    International Nuclear Information System (INIS)

    Gomez-Briceno, Dolores; Lapena, Jesus; Castano, M. Luisa; Blazquez, Fernando

    2002-01-01

    Full text: Cracking due to PWSCC in PWR CRDM nozzles and other VHP nozzles fabricated from Alloy 600 is not a new issue. In 1991, a leak was discovered on one CRDM nozzle at Bugey 3 PWR plant in France. The cause of the cracking was identified as primary water stress corrosion cracking. From then, similar cracks have been found in other European and USA PWR plants. The cracks were predominantly axial in orientation and it was accepted that CRDM nozzles and weld cracking in PWR was not a immediate safety concern. However, this consideration has to be reassessed in light of the recent identification of circumferential cracking in CRDM nozzles at Oconee Nuclear Station Unit 2 and 3 along with axial cracking in the Alloy 182 J-groove welds at these two units and at Oconee Nuclear Station 1 and Arkansas Nuclear One Unit 1. Alloy 600 susceptibility in primary water has received an enormous research effort for many years since the Alloy 600 steam generators tube degradation started. A significant amount of information is available to characterise the susceptibility of Alloy 600. However, Alloy 600 susceptibility is strongly dependent on the heat thermomechanical history and both the crack initiation time and the crack growth rate data obtained from representative materials of the VHP nozzles seem to be necessary for the structural integrity assessment of cracking nozzles. An extensive experimental program has been performed at CIEMAT, to study the behaviour of Alloy 600 VHP nozzles in PWR primary conditions. Crack initiation and crack propagation tests have been performed using different types of products (forged bar, tube, plate and steam generator tubing). Long duration crack initiation tests have been carried out, at 330 deg. C and 360 deg. C in water and at 400 deg. C in steam, using ten Alloy 600 heats with yield strength ranging from 291 MPa to 489 MPa. The influence of several parameters (grain boundary carbide distribution, grain size and yield strength) on crack

  16. PWR primary system chemistry control during hot functional testing

    International Nuclear Information System (INIS)

    Reid, Richard D.; Little, Michael J.

    2014-01-01

    Hot Functional Testing (HFT) involves a number of pre-operational exercises performed to confirm the operability of plant systems at conditions expected during both normal and off-normal operation of a pressurized water reactor (PWR), including operability of safety systems. While the primary purposes of HFT are to demonstrate operability of plant systems and satisfy regulatory requirements, chemistry control during HFT is important to long-term integrity and performance of plant systems. Specifically, HFT is the first time plant equipment is exposed to high temperature water and the chemistry maintained during HFT can impact the passivation layers that form on wetted surfaces and long-term release of metals from these surfaces. Metals released from the inner surfaces of steam generator tubing and reactor coolant loop piping become activated in the core and can redeposit on ex-core surfaces. Because HFT is performed before fuel is loaded in the core, HFT provides an opportunity to produce a passive layer on primary surfaces that is free of activated corrosion products, resistant to metals release during subsequent plant operation, and also resistant to incorporation of activated corrosion products (once fuel is loaded in the core). Thus, maintaining desirable primary chemistry control during HFT is important for source term management, minimization of future shutdown activity releases, minimization of dose rates, and asset preservation. This paper presents an overview of passive film formation in the austenitic stainless steel and high nickel alloys that make up the majority of the primary circuit in advanced PWR designs. Based on this information, a summary is provided of the effects on passive film formation of key chemistry parameters that may be controlled during HFT. (author)

  17. Behaviour of fission products in PWR primary coolant and defected fuel rods evaluation

    International Nuclear Information System (INIS)

    Bourgeois, P.; Stora, J.P.

    1979-01-01

    The activity surveillance of the PWR primary coolant by γ spectometry gives some informations on fuel failures. The activity of different nuclides e.g. Xenons, Kryptons, Iodines, can be correlated with the number of the defected fuel rods. Therefore the precharacterization with eventually a prelocalization of the related fuel assemblies direct the sipping-test and allows a saving of time during refueling. A model is proposed to calculate the number of the defected rods from the activity measurements of the primary coolant. A semi-empirical model of the release of the fission products has been built from the activity measurements of the primary coolant in a 900 MWe PWR. This model allows to calculate the number of the defected rods and also a typical parameter of the mean damage. Fission product release is described by three stages: release from uranium dioxide, transport across the gas gap and behaviour in the primary coolant. The model of release from the oxide considers a diffusion process in the grains with trapping. The release then occurs either directly to free surfaces or with a delay due to a transit into closed porosity of the oxide. The amount released is the same for iodine and rare gas. With the gas gap transit is associated a transport time and a probability of trapping for the iodines. In the primary coolant the purification and the radioactive decay are considered. (orig.)

  18. Shutdown Chemistry Process Development for PWR Primary System

    Energy Technology Data Exchange (ETDEWEB)

    Sung, K.B. [Korea Electric Power Research Institute, Taejeon (Korea, Republic of)

    1997-12-31

    This study report presents the shutdown chemistry of PWR primary system to reduce and remove the radioactive corrosion products which were deposited on the nuclear fuel rods surface and the outside of core like steam generator channel head, RCS pipings etc. The major research results are the follows ; the deposition radioactive mechanism of corrosion products, the radiochemical composition, the condition of coolant chemistry to promote the dissolution of radioactive cobalt and nickel ferrite, the control method of dissolved hydrogen concentration in the coolant by the mechanical and chemical methods. The another part of study is to investigate the removal characteristics of corrosion product ions and particles by the demineralization system to suggest the method which the system could be operate effectively in shut-down purification period. (author). 19 refs., 25 figs., 48 tabs.

  19. Experimental studies of PWR primary piping under loca

    International Nuclear Information System (INIS)

    Caumette, Pierre; Garcia, J.L.

    1980-07-01

    The experimental program performed on AQUITAINE II facility is directed to study the mechanical behavior of primary PWR pipes and the forces exerted on the neighbouring structures as a consequence of a breach opening. It has been developed in the form of a quadripartite agreement between the Commissariat a l'Energie Atomique, Framatome, Electricite de France and Westinghouse. Some forty tests have been carried out with different pipe configurations (straight tube, elbow, S- or U-shaped tube) and different break types (single or double guillotine). The following aspects are investigated: - the dynamic behavior of the pipe and in particular the formation of a plastic hinge at the restraint; - the impact function of a pipe or an energy-absorbing bumper; - the lateral stability of both ends of a pipe, after a double-guillotine break [fr

  20. Tendency of nuclear pumps for PWR primary system

    International Nuclear Information System (INIS)

    Shibata, Takeshi

    1976-01-01

    At present, large PWR power stations of more than 1,000 MW are successively constructed, and the pumps used there have become large. The progress and tendency of the technical development of main pumps in primary system are described. The increase of the capacity of power stations is accomplished by increasing the circulating coolant quantity per loop or the number of loops. Same standard primary coolant pumps are employed in the plants from 500 to 1,100 MW. The type of primary coolant pumps changed from canned type to shaft seal type, and the advantages of the shaft seal type are cheap production cost, high efficiency, and the easy utilization of inertia force. The bearings and shaft seals are thermally insulated from primary coolant. As for auxiliary pumps, reciprocating filling-up pumps and centrifugal high pressure injection pumps are used for 500 MW plants, but only centrifugal pumps are used for both purposes in 800 MW plants, and in 1,100 MW plants, the pumps of both types for separate purposes and centrifugal pumps for combined purposes are installed. Horizontal or vertical pumps of same type are used as containment vessel-spraying pumps and excess heat-eliminating pumps. The type of boric acid pumps changed from canned type to mechanical seal type. (Kako, I.)

  1. Contamination of a green algae (Scenedesmus obliquus) from fresh water by radionuclides typical of PWR effluents: culture in a turbidostat

    International Nuclear Information System (INIS)

    Sombre, L.; Carraro, S.; Myttenaere, C.

    1987-01-01

    The fixation of radioactive polluents, typical for PWR effluents, by the green soft water algae Scenedesmus obliquus is studied by means of a continuous culture method and in controlled conditions (turbidostat). Transfer factors are obtained. The elimination of radiocesium occurs in two distinct phases characterized respectively by a short biological period of less than one hour and a long period of the order of one day. The photosynthetic metabolism of the algae accounts for 25% of the decorporation. (Author)

  2. 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

  3. Qualitative analysis of the maintenance politics of the systems of a typical PWR by artificial neural networks

    International Nuclear Information System (INIS)

    Lourenco, Victor Hugo Moreno

    2010-02-01

    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)

  4. PKL/K9, Refill and Reflood Experiment in a Simulated PWR Primary System (PKL)

    International Nuclear Information System (INIS)

    1981-01-01

    1 - Description of test facility: PKL-facility simulates the essential primary system components of a typical West German 1300 PWR with regard to their thermohydraulic behaviour. The facility essentially consists of the pressure vessel with the heated bundle, the downcomer simulator, the primary loops with the components steam generator and pump simulator, the injection devices, the break geometry simulator, as well as the separators connected thereto, and the test containment to maintain a back-pressure at the location of break which is expected to be typical for emergency conditions. The number of heater rods and the cross-sections of the testing plant are on a reduced scale 1:134 in comparison with a typical German PWR. The elevations and locations are essentially full scale. Pressure vessel: The space between the pressure vessel and the inner core casing is sealed from the core region and the upper and lower plenum and connected with the upper plenum only by a pressure equalization line. The rod bundle surrounded by the inner core casing consists of 340 rods, 337 of which are indirect electrically heated. The test bundle cross-section as well as a heater element with the measuring elevations, the original-KWU-spacers and the axial power profile (7 power steps) are described. Downcomer: The downcomer is simulated by the downcomer nozzle region and the downcomer U-tube. The cold leg injection takes place both directly in the downcomer nozzle region and in the lines of t he intact single and double loop near to the downcomer nozzle region. A cylindrical insertion and repulsing metal sheets are installed in the downcomer nozzle region in order to avoid the emergency injection points into the broken loop. 2 - Description of test: Test K 9 out of a series PKL-IB was conducted on May 30, 1979 by Kraftwerk Union (KWU) at Erlangen (Germany). The objective of the integral cold leg injection test K 9 (double-ended 200%-break) was to investigate after a LOCA the refill and

  5. Dose rate determining factors of PWR primary water

    International Nuclear Information System (INIS)

    Terachi, Takumi; Kuge, Toshiharu; Nakano, Nobuo

    2014-01-01

    The relationship between dose rate trends and water chemistry has been studied to clarify the determining factors on the dose rates. Therefore dose rate trends and water chemistry of 11 PWR plants of KEPCO (Kansai Electric Power Co., Inc.) were summarized. It is indicated that the chemical composition of the oxide film, behaviour of corrosion products and Co-58/Co-60 ratio in the primary system have effected dose rate trends based on plant operation experiences for over 40 years. According to plant operation experiences, the amount of Co-58 has been decreasing with the increasing duration of SG (Steam Generator) usage. It is indicated that the stable oxide film formation on the inner surface of SG tubing, is a major beneficial factor for radiation sources reduction. On the other hand, the reduction of the amount of Co-60 for the long term has been not clearly observed especially in particular high dose plants. The primary water parameters imply that considering release and purification balance on Co-59 is important to prevent accumulation of source term in primary water. In addition, the effect of zinc injection, which relates to the chemical composition of oxide film, was also assessed. As the results, the amount of radioactive Co has been clearly decreased. The decreasing trend seems to correlate to the half-life of Co-60, because it is considered that the injected zinc prevents the uptake of radioactive Co into the oxide film on the inner surface of the components and piping. In this paper, the influence of water chemistry and the replacement experiences of materials on the dose rates were discussed. (author)

  6. Development of forging technology for PWR primary piping

    International Nuclear Information System (INIS)

    Morin, F.; Badeau, J.P.; Lambs, R.

    1996-01-01

    The purpose of this presentation is to give information on the changes in the design and manufacture of Primary Piping for electronuclear boilers of the Pressurized Water Reactor type (PWR) which has resulted in the making of one-piece forged lines including stub pipes and arcs. The optimization of these items is aimed at improving the life of the new power stations as well as guaranteeing their safety, while reducing inspection and maintenance requirements in service. The demonstration of the manufacturing feasibility has just been completed. It has taken material form in the installation, on the CIVAUX 1 section, of the first one-piece cold leg in the world. It will shortly be followed by the installation on the CIVAUX 2 section of a complete loop of bent forged pipes. Therefore, this new know-how is going to be incorporated in the French Rules (RCC-M) and can be directly taken into consideration both in the next work to be done and in the design and definition of a future nuclear reactor

  7. Comparison of computational performance of GA and PSO optimization techniques when designing similar systems - Typical PWR core case

    Energy Technology Data Exchange (ETDEWEB)

    Souza Lima, Carlos A. [Instituto de Engenharia Nuclear - Divisao de Reatores/PPGIEN, Rua Helio de Almeida 75, Cidade Universitaria - Ilha do Fundao, P.O. Box: 68550 - Zip Code: 21941-972, Rio de Janeiro (Brazil); Instituto Politecnico, Universidade do Estado do Rio de Janeiro, Pos-Graduacao em Modelagem Computacional, Rua Alberto Rangel - s/n, Vila Nova, Nova Friburgo, Zip Code: 28630-050, Nova Friburgo (Brazil); Lapa, Celso Marcelo F.; Pereira, Claudio Marcio do N.A. [Instituto de Engenharia Nuclear - Divisao de Reatores/PPGIEN, Rua Helio de Almeida 75, Cidade Universitaria - Ilha do Fundao, P.O. Box: 68550 - Zip Code: 21941-972, Rio de Janeiro (Brazil); Instituto Nacional de Ciencia e Tecnologia de Reatores Nucleares Inovadores (INCT) (Brazil); Cunha, Joao J. da [Eletronuclear Eletrobras Termonuclear - Gerencia de Analise de Seguranca Nuclear, Rua da Candelaria, 65, 7 andar. Centro, Zip Code: 20091-906, Rio de Janeiro (Brazil); Alvim, Antonio Carlos M. [Universidade Federal do Rio de Janeiro, COPPE/Nuclear, Cidade Universitaria - Ilha do Fundao s/n, P.O.Box 68509 - Zip Code: 21945-970, Rio de Janeiro (Brazil); Instituto Nacional de Ciencia e Tecnologia de Reatores Nucleares Inovadores (INCT) (Brazil)

    2011-06-15

    Research highlights: > Performance of PSO and GA techniques applied to similar system design. > This work uses ANGRA1 (two loop PWR) core as a prototype. > Results indicate that PSO technique is more adequate than GA to solve this kind of problem. - Abstract: This paper compares the performance of two optimization techniques, particle swarm optimization (PSO) and genetic algorithm (GA) applied to the design a typical reduced scale two loop Pressurized Water Reactor (PWR) core, at full power in single phase forced circulation flow. This comparison aims at analyzing the performance in reaching the global optimum, considering that both heuristics are based on population search methods, that is, methods whose population (candidate solution set) evolve from one generation to the next using a combination of deterministic and probabilistic rules. The simulated PWR, similar to ANGRA 1 power plant, was used as a case example to compare the performance of PSO and GA. Results from simulations indicated that PSO is more adequate to solve this kind of problem.

  8. Comparison of computational performance of GA and PSO optimization techniques when designing similar systems - Typical PWR core case

    International Nuclear Information System (INIS)

    Souza Lima, Carlos A.; Lapa, Celso Marcelo F.; Pereira, Claudio Marcio do N.A.; Cunha, Joao J. da; Alvim, Antonio Carlos M.

    2011-01-01

    Research highlights: → Performance of PSO and GA techniques applied to similar system design. → This work uses ANGRA1 (two loop PWR) core as a prototype. → Results indicate that PSO technique is more adequate than GA to solve this kind of problem. - Abstract: This paper compares the performance of two optimization techniques, particle swarm optimization (PSO) and genetic algorithm (GA) applied to the design a typical reduced scale two loop Pressurized Water Reactor (PWR) core, at full power in single phase forced circulation flow. This comparison aims at analyzing the performance in reaching the global optimum, considering that both heuristics are based on population search methods, that is, methods whose population (candidate solution set) evolve from one generation to the next using a combination of deterministic and probabilistic rules. The simulated PWR, similar to ANGRA 1 power plant, was used as a case example to compare the performance of PSO and GA. Results from simulations indicated that PSO is more adequate to solve this kind of problem.

  9. The impact of steam generator replacement on PWR primary system contamination

    International Nuclear Information System (INIS)

    Dacquait, F.; Marteau, H.; Guinard, L.; Ranchoux, G.; Taunier, S.; Wintergerst, M.; Bretelle, J.L.; Rocher, A.

    2010-01-01

    This paper analyses the impact of Steam Generator Replacement (SGR) on PWR primary circuit contamination. It presents a comparison of the activities deposited inside the primary system and released during refuelling outages after SGR with three different SG tube alloys (600, 690 and 800) and different SG tube manufacturing processes. A SGR has a great impact on the primary system contamination. After SGR, whatever the SG tube material is, the typical variations are the following: The 58 Co contamination increases for 1 to 3 cycles, and then decreases to very low levels in some cases, mainly depending on the manufacturing process of the replacement SG tubes; The 60 Co Co contamination tends to decrease on the primary coolant pipes and increases by a lower rate on the new SG tubes. This analysis highlights the importance on contamination levels after SGR of both the corrosion product deposits on the primary surfaces before SGR and the surface finish of the SG tubes related to their manufacturing process. (author)

  10. Modelling of the local chemistry in stagnant areas in the PWR primary circuit

    International Nuclear Information System (INIS)

    Reid, Rick; Fruzzetti, Keith; Ahluwalia, Al; Summe, Alex; Dame, Cecile; Schmitt, Kyle

    2014-01-01

    MRP-236 demonstrated a correlation between stagnant or low flow conditions and stress corrosion cracking (SCC) of stainless steel components in the PWR primary system. Of the approximately 140 SCC events documented (affecting 15 different components), 83% involved stagnant or low flow conditions that were likely to be associated with chemical environments different from the well mixed bulk coolant. The chemistry in such locations is typically not monitored, and sampling is difficult or impossible. Actions to improve chemistry in regions of low or no coolant flow, such as flushing, cycling of components and imposition of more stringent make up water chemistry controls affect both operational costs and outage schedules. Similarly, design changes to improve flow in affected areas are costly or impracticable. Improving the understanding of the factors controlling chemistry in such areas and development of the capability to predict typical and worst case conditions will allow an informed assessment of procedural actions and/or design changes to improve local chemistry and thereby reduce SCC susceptibility. A project was undertaken to develop a model to predict local chemistry conditions in stagnant locations. The model comprises the iterative application of the EPRI MULTEQ solution chemistry equilibrium code and standard thermodynamic relationships to predict local chemistry conditions considered likely to have been present at the surfaces of components when SCC was initiated. The starting chemistry conditions are based on PWR primary system chemistry from different plant maneuvers (e.g., startup and shutdown conditions). The model was applied to three example components where SCC has occurred in the field. The selected components were: control rod drive mechanism canopy seals; valve drain lines; and reactor vessel o-ring leak-off lines. This paper provides a summary of the model and predicted local chemistry conditions that develop for the three example component as a

  11. Evaluation of ASME code flaw analysis procedure using the influence function method for application to PWR primary piping

    International Nuclear Information System (INIS)

    Hong, S.Y.; Yeater, M.L.

    1985-01-01

    This paper discusses stress intensity factor calculations and fatigue analysis for a PWR primary coolant piping system. The influence function method is applied to evaluate ASME Code Section XI Appendix A ''analysis of flaw indication'' for the application to a PWR primary piping. Results of the analysis are discussed in detail. (orig.)

  12. Studies of fuel loading pattern optimization for a typical pressurized water reactor (PWR) using improved pivot particle swarm method

    International Nuclear Information System (INIS)

    Liu, Shichang; Cai, Jiejin

    2012-01-01

    Highlights: ► The mathematical model of loading pattern problems for PWR has been established. ► IPPSO was integrated with ‘donjon’ and ‘dragon’ into fuel arrangement optimizing code. ► The novel method showed highly efficiency for the LP problems. ► The core effective multiplication factor increases by about 10% in simulation cases. ► The power peaking factor decreases by about 0.6% in simulation cases. -- Abstract: An in-core fuel reload design tool using the improved pivot particle swarm method was developed for the loading pattern optimization problems in a typical PWR, such as Daya Bay Nuclear Power Plant. The discrete, multi-objective improved pivot particle swarm optimization, was integrated with the in-core physics calculation code ‘donjon’ based on finite element method, and assemblies’ group constant calculation code ‘dragon’, composing the optimization code for fuel arrangement. The codes of both ‘donjon’ and ‘dragon’ were programmed by Institute of Nuclear Engineering of Polytechnique Montréal, Canada. This optimization code was aiming to maximize the core effective multiplication factor (Keff), while keeping the local power peaking factor (Ppf) lower than a predetermined value to maintain fuel integrity. At last, the code was applied to the first cycle loading of Daya Bay Nuclear Power Plant. The result showed that, compared with the reference loading pattern design, the core effective multiplication factor increased by 9.6%, while the power peaking factor decreased by 0.6%, meeting the safety requirement.

  13. Possibility of Localized Corrosion of PWR primary side materials in oxidative decontamination condition

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Yoon; Jung, Jun Young; Won, Huijun; Kim, Seon Byeong; Choi, Wangkyu; Moon, Jeikwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Primary circuit of a PWR (radionuclides uptake in the inner oxide layer and oxide/metal interface occurred inevitably. Therefore, it is necessary to remove the inner oxide layer as well as the outer oxide layer to achieve excellent decontamination effects. It is known that the outer oxide layers are typically composed of Fe{sub 3}O{sub 4} and NiFe{sub 2}O{sub 4}. On the other hand, the inner oxide layers are composed of Cr{sub 2}O{sub 3}, (Ni{sub 1-x}Ni{sub x})(Cr{sub 1-y}Fe{sub y}){sub 2}O{sub 4}, and FeCr{sub 2}O{sub 4}. Because of chromium in the trivalent oxidation state which is difficult to dissolve, the oxide layer has an excellent protectiveness and is hard to decontaminate. For the dissolution of chromium-rich oxide, there have been developed an alkaline permanganate (AP) or nitric permanganate (NP). A disadvantage of the AP process is the generation of a large volume of secondary waste. On the other hand, NP process is highly incompatible to the corrosion of the structure materials. In this study as a part of developing decontamination process, we investigated the corrosion behavior of the structure materials such as Inconel-600 and type 304 stainless steel in NP and AP oxidative decontamination conditions for the safe use of an oxidative phase in PWR system decontamination. The corrosion behavior was analyzed through the potential-pH equilibrium for the system of Cr-H{sub 2}O / Mn-H{sub 2}O at 90 .deg. C and potentiodynamic polarization in a typical AP and NP solution were evaluated. The AP or NP treated specimen surface was observed using an optical microscope and scanning electron microscopy (SEM) for an evaluation of the localized corrosion. The possibility of localized corrosion of PWR primary side materials under oxidative decontamination condition was evaluated using a potentiodynamic polarization technique, observation of localized corrosion morphology, and consideration of potential-pH diagrams at 90 .deg. C. From the results of these tests, we

  14. Investigation of chloride-release of nuclear grade resin in PWR primary system coolant

    International Nuclear Information System (INIS)

    Cao Xiaoning; Li Yunde; Li Jinghong; Lin Fangliang

    1997-01-01

    A new preparation technique is developed for making the low-chloride nuclear-grade resin by commercial resin. The chloride remained in nuclear grade resin may release to PWR primary coolant. The amount of released chloride is depended on the concentration of boron, lithium, other anion impurities, and remained chloride concentration in resin

  15. Simulation of a severe accident at a typical PWR due to break of a hot leg ECCS line using MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung Min; Sabundjian, Gaianê, E-mail: smlee@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2017-11-01

    The aim of this work was to simulate a severe accident at a typical PWR caused by break in Emergency Core Cooling System (ECCS) line of a hot leg using the MELCOR code. The nodalization of this typical PWR was elaborated by the Global Research for Safety (GRS) and provided to the CNEN for analysis of the severe accidents at the Angra 2, which is similar to that PWR. Although both of them are not identical the results obtained for that typical PWR may be valuable because of the lack of officially published calculation for Angra 2. Relevant parameters such as pressure, temperature and water level in various control volumes after the break in the hot leg were calculated as well as degree of core degradation and hydrogen concentration in containment. The result obtained in this work could be considered satisfactory in the sense that the physical phenomena reproduced by the simulation were in general very reasonable, and most of the events occurred within acceptable time intervals. However, the uncertainty analysis was not carried out in this work. Furthermore, this scenario could be used as a base for the study of the effectiveness of some preventive or/and mitigating measures of Severe Accident Management (SAMG) by adding associated conditions for each measure in its input. (author)

  16. Simulation of a severe accident at a typical PWR due to break of a hot leg ECCS line using MELCOR code

    International Nuclear Information System (INIS)

    Lee, Seung Min; Sabundjian, Gaianê

    2017-01-01

    The aim of this work was to simulate a severe accident at a typical PWR caused by break in Emergency Core Cooling System (ECCS) line of a hot leg using the MELCOR code. The nodalization of this typical PWR was elaborated by the Global Research for Safety (GRS) and provided to the CNEN for analysis of the severe accidents at the Angra 2, which is similar to that PWR. Although both of them are not identical the results obtained for that typical PWR may be valuable because of the lack of officially published calculation for Angra 2. Relevant parameters such as pressure, temperature and water level in various control volumes after the break in the hot leg were calculated as well as degree of core degradation and hydrogen concentration in containment. The result obtained in this work could be considered satisfactory in the sense that the physical phenomena reproduced by the simulation were in general very reasonable, and most of the events occurred within acceptable time intervals. However, the uncertainty analysis was not carried out in this work. Furthermore, this scenario could be used as a base for the study of the effectiveness of some preventive or/and mitigating measures of Severe Accident Management (SAMG) by adding associated conditions for each measure in its input. (author)

  17. Study on B-10 consumption of PWR primary coolant during normal operation

    International Nuclear Information System (INIS)

    Liang, C.H.

    1994-01-01

    B-10 consumption under PWR primary coolant conditions has been analyzed. The result indicates its time-dependent change reacting with neutron in the normal operation. In this work, neutron energy assumed to be 4 eV; thermal neutron flux is in the range of 3 x 10 13 to 3 x 10 14 n/sec - cm 2 and the time of cycling of the primary coolant through the RCS is 8 sec. and its retention time in the core region is about 1 sec. Under this condition investigated, B-10 consumption is less than 5% at 3 x 10 13 n/sec - cm 2 thermal neutron flux, and closes to 27% at 3 x 10 14 n/sec - cm 2 by calculation at the 16th month of continuous operation. The effect of B-10 consumption on PWR primary water chemistry is also investigated. (author). 1 fig., 2 tabs., 4 refs

  18. Alternative water chemistry for the primary loop of PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Henzel, N [Siemens AG Unternehmensbereich KWU, Erlangen (Germany)

    1997-02-01

    Advanced fuel element concepts (longer cycles, higher burnup, increased rod power) call for more reactivity binding capacity and, moreover, might produce higher void fractions, particularly in the hot channel. Thus, on the one hand, more alcalizing agent is needed to maintain a high coolant pH according to the approved ``modified boron-lithium mode of operation`` in the presence of more boric acid (chemical shim); on the other hand, increasing enrichment of coolant constituents due to local boiling (higher void fraction), which must not result in accelerated corrosion of fuel cladding and structural materials, imposes enhanced requirements on both, materials technology and water chemistry. At present, the use of boric acid enriched in B10 (the isotope effective in terms of reactivity control) appears to advantageously compromise in capturing more neutrons with less total boron while maintaining or even slightly reducing lithium concentrations at the same time. There is no feasible alternative for boric acid used as the chemical shim and for hydrogen gas as the reducing agent used to suppress oxygen formation by water radiolysis. Systematic screening as performed in phase 1 of a recent project proved potassium hydroxide to be the only potential candidate to favourably replace lithium 7 hydroxide as an alcalizing agent. Unfortunately, the results of pertinent comparative corrosion tests are not unambiguous, and available operational experience with potassium hydroxide in WWER plants is not readily applicable to western world-type PWR plants. Therefore, a switch-over from lithium to potassium can be envisaged only subsequent to a comprehensive qualification program which is planned to be the objective of phase 2 of the project. This program should also comprise zinc addition tests in order to confirm the alleged positive impact of this element on corrosion rates and activity buildup. (Abstract Truncated)

  19. Evaluation of CRUDTRAN code to predict transport of corrosion products and radioactivity in the PWR primary coolant system

    International Nuclear Information System (INIS)

    Lee, C.B.

    2002-01-01

    CRUDTRAN code is to predict transport of the corrosion products and their radio-activated nuclides such as cobalt-58 and cobalt-60 in the PWR primary coolant system. In CRUDTRAN code the PWR primary circuit is divided into three principal sections such as the core, the coolant and the steam generator. The main driving force for corrosion product transport in the PWR primary coolant comes from coolant temperature change throughout the system and a subsequent change in corrosion product solubility. As the coolant temperature changes around the PWR primary circuit, saturation status of the corrosion products in the coolant also changes such that under-saturation in steam generator and super-saturation in the core. CRUDTRAN code was evaluated by comparison with the results of the in-reactor loop tests simulating the PWR primary coolant system and PWR plant data. It showed that CRUDTRAN could predict variations of cobalt-58 and cobalt-60 radioactivity with time, plant cycle and coolant chemistry in the PWR plant. (author)

  20. Techniques for Primary-to-Secondary Leak Monitoring in PWR Plants

    International Nuclear Information System (INIS)

    Sohn, Wook; Chi, Jun Hwa; Kang, Duck Won; Tae, Jeong Woo

    2006-01-01

    Historically, corrosion and mechanical damage have made steam generator tubes in PWR plants see various types of degradation from both the primary and secondary sides of the tubes. Since the tube degradation can lead to through-wall failure, the plant personnel should make efforts to prevent the failure. One of such preventive efforts is to monitor primary-to-secondary leakage (PSL) that usually precedes the tube rupture. Thus the objective of PSL monitoring is to make operators to determine when to shutdown the plant in order to minimize the likelihood of propagation of leaks to tube rupture under normal and faulted conditions This paper addresses briefly the status of techniques for PSL monitoring used in PWR plants

  1. Recent bibliography on analytical and sampling problems of a PWR primary coolant Pt. 1

    International Nuclear Information System (INIS)

    Illy, H.

    1981-12-01

    The first bibliography on analytical and sampling problems of a PWR primary coolant (KFKI Report-1980-48) was published in 1980 and it covered the literature published in the previous 8-10 years. The present supplement reviews the subsequent literature up till December 1981. It also includes some references overlooked in the first volume. The serial numbers are continued from the first bibliography. (author)

  2. Hard alloys testing-machine for values of PWR primary coolant circuits

    International Nuclear Information System (INIS)

    Campan, J.L.; Sauze, A.

    1980-01-01

    Testing of valve parts or material used in valve fabrication and particularly seizing conditions in friction of plane surfaces coated with hard alloys of the type stellite. The testing equipment called Marguerite is composed of a hot pressurized water loop in conditions similar to PWR primary coolant circuits (320 0 C, 150 bars) and a testing-machine with measuring instruments. Testing conditions and samples are described [fr

  3. Analytical and sampling problems in primary coolant circuits of PWR-type reactors

    International Nuclear Information System (INIS)

    Illy, H.

    1980-10-01

    Details of recent analytical methods on the analysis and sampling of a PWR primary coolant are given in the order as follows: sampling and preparation; analysis of the gases dissolved in the water; monitoring of radiating substances; checking of boric acid concentration which controls the reactivity. The bibliography of this work and directions for its use are published in a separate report: KFKI-80-48 (1980). (author)

  4. Primary water chemistry improvement for radiation exposure reduction at Japanese PWR Plants

    Energy Technology Data Exchange (ETDEWEB)

    Nishizawa, Eiichi [Omiya Technical Institute, Saitama-ken (Japan)

    1995-03-01

    Radiation exposure during the refueling outages at Japanese Pressurized Water Reactor (PWR) Plants has been gradually decreased through continuous efforts keeping the radiation dose rates at relatively low level. The improvement of primary water chemistry in respect to reduction of the radiation sources appears as one of the most important contributions to the achieved results and can be classified by the plant operation conditions as follows

  5. Experiments for simulating a great leak in the primary coolant circuit of a PWR type reactor

    International Nuclear Information System (INIS)

    Liebig, E.

    1977-01-01

    A loss of coolant accident is to be simulated on a high pressure test rig. The accident is initiated by an externally induced rupture of a pair of rupture-disks installed in a coolant ejection device. Several problems of simulating leaks in the primary coolant circuit of PWR type reactors are dealt with. The selection of appropriate rupture-disks for such experiments is described

  6. Recent bibliography on analytical and sampling problems of a PWR primary coolant Suppl. 4

    International Nuclear Information System (INIS)

    Illy, H.

    1986-09-01

    The 4th supplement of a bibliographical series comprising the analytical and sampling problems of the primary coolant of PWR type reactors covers the literature from 1985 up to July 1986 (220 items). References are listed according to the following topics: boric acid; chloride, chlorine; general; hydrogen isotopes; iodine; iodide; noble gases; oxygen; other elements; radiation monitoring; reactor safety; sampling; water chemistry. (V.N.)

  7. An Overview of the EPRI PWR Primary Chemistry Program

    International Nuclear Information System (INIS)

    Perkins, David; Fruzzetti, Keith; Haas, Carey; Wells, Dan

    2012-09-01

    Primary chemistry controls continue to evolve, impacting long term equipment reliability goals, optimized core designs, and radiation dose management practices. Chemistry initiatives include increased primary system pH (T) , zinc injection, and optimization of primary system hydrogen concentration. Nevertheless, utilities are faced with ever changing challenges as fuel vendors continue to optimize core power densities coupled with longer operating cycles and material replacement efforts. These challenges must be collaboratively addressed by the plant chemists, engineers, and operators. Operational chemistry has changed dramatically over the years with increased primary pH (T) programs requiring some utilities to operate with up to 6 ppm lithium or slightly higher. Coupled with primary pH (T) program optimization, are ongoing EPRI research efforts attempting to develop an optimized hydrogen control program balancing material issues associated with primary water stress corrosion cracking (PWSCC) crack growth rate against fuel concerns associated with increased hydrogen concentrations. One of the most significant primary chemistry changes that effectively balances the demands of materials, fuels, chemistry and dose management strategies is zinc injection into the primary coolant. Since 1994 when Farley initiated zinc injection, zinc injection has been successfully injected at over 70 pressurized water reactors world-wide. Combining operational chemistry with shutdown chemistry controls provides the plant chemist with a technically based and balanced approach to fuel and material integrity as well as dose management strategies. Shutdown chemistry has continually evolved since the 1970's when the chemist was primarily concerned with fission products. Now the chemist must manage corrosion product release, and support Outage Management and Radiation Protection through the performance of a controlled shutdown. In part, this change was driven as plant materials evolved

  8. Metallurgical and mechanical parameters controlling alloy 718 stress corrosion cracking resistance in PWR primary water

    International Nuclear Information System (INIS)

    Deleume, J.

    2007-11-01

    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]. δ 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. 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)

  10. The empirical intensity of PWR primary coolant pumps failure and repair

    International Nuclear Information System (INIS)

    Milivojevicj, S.; Riznicj, J.

    1988-01-01

    The wealth of operating experience concerning PWR type and nuclear reactors that has been regularly monitored and systematically processes since 1971, enabled an analysis of the PWR primary coolant pumps operation. Failure intensity α and repair intensity μ of the pump during its working life were calculated, as these values are necessary in order to determine the reliability and availability of the pump as the basis for analyzing its effect on the safety and efficiency of the nuclear power plant. The trend of failure intensity α follows the theoretically expected changes in α over time, and this is around 10 -5 in the majority of life-time. Repair intensity μ indicates a slow rise during life-time, i.e. its faster return to operation. (author).7 refs.; 5 figs

  11. DETERMINATION OF THE 129I IN PRIMARY COOLANT OF PWR

    Directory of Open Access Journals (Sweden)

    KE CHON CHOI

    2013-02-01

    In this report, the effect of the boron content in a pressurized-water reactor primary coolant on the separation process of 129I was examined, as was the effect of 3H on the measurement of the activity of iodine. As a result, no influence of the boron content and of the simultaneous 3H presence was found with activity concentrations of 3H lower than 50 Bq/mL, and with a boron concentration of less than 2,000 μg/mL.

  12. Solubility of simulated PWR primary circuit corrosion products

    International Nuclear Information System (INIS)

    Kunig, R.H.; Sandler, Y.L.

    1986-08-01

    The solubility behavior of non-stoichiometric nickel ferrites, nickel-cobalt ferrites, and magnetite, as model substances for the corrosion products (''crud'') formed in nuclear pressurized water reactors, was studied in a flow system in aqueous solutions of lithium hydroxide, boric acid, and hydrogen with pH, temperature, and hydrogen concentrations as parameters. Below the temperature region of 300 to 330 0 C, at hydrogen concentrations of 25 to 40 cm 3 /kg H 2 O as used during reactor operation, the solubility of nickel-cobalt ferrite is the same as that of Ni and Co/sub x/Fe/sub 3-x/O 4 (x 3 /kg of hydrogen, the equilibrium iron and nickel solubilities increase congruently down to about 100 0 C, in a manner consistent with the solubility of Fe 3 O 4 , but sharply decline at lower temperatures, apparently due to formation of a borated layer. A cooldown experiment on a time scale of a typical Westinghouse reactor shutdown, as well as static experiments carried out on various ferrite samples at 60 0 C show that after addition of oxygen or peroxide evolution of nickel (and possibly cobalt) above the equilibrium solubility in hydrogen depends on the presence of dissociation products prior to oxidation. Thermodynamic calculations of various reduction and oxidative decomposition reactions for stoichiometric and non-stoichiometric nickel ferrite and cobalt ferrite are presented. Their significance to evolutions of nickel and cobalt on reactor shutdown is discussed. 30 refs., 38 figs., 34 tabs

  13. Stress corrosion cracking of nickel base alloys in PWR primary water

    International Nuclear Information System (INIS)

    Guerre, C.; Chaumun, E.; Crepin, J.; De Curieres, I.; Duhamel, C.; Heripre, E.; Herms, E.; Laghoutaris, P.; Molins, R.; Sennour, M.; Vaillant, F.

    2013-01-01

    Stress corrosion cracking (SCC) of nickel base alloys and associated weld metals in primary water is one of the major concerns for pressurized water reactors (PWR). Since the 90's, highly cold-worked stainless steels (non-sensitized) were also found to be susceptible to SCC in PWR primary water ([1], [2], [3]). In the context of the life extension of pressurized water reactors, laboratory studies are performed in order to evaluate the SCC behaviour of components made of nickel base alloys and of stainless steels. Some examples of these laboratory studies performed at CEA will be given in the talk. This presentation deals with both initiation and propagation of stress corrosion cracks. The aims of these studies is, on one hand, to obtain more data regarding initiation time or crack growth rate and, one the other hand, to improve our knowledge of the SCC mechanisms. The aim of these approaches is to model SCC and to predict components life duration. Crack growth rate (CGR) tests on Alloy 82 with and without post weld heat treatment are performed in PWR primary water (Figure 1). The heat treatment seems to be highly beneficial by decreasing the CGR. This result could be explained by the effect of thermal treatment on the grain boundary nano-scopic precipitation in Alloy 82 [4]. The susceptibility to SCC of cold worked austenitic stainless steels is also studied. It is shown that for a given cold-working procedure, SCC susceptibility increases with increasing cold-work ([2], [5]). Despite the fact that the SCC behaviour of Alloy 600 has been widely studied for many years, recent laboratory experiments and analysis ([6], [7], [8]) showed that oxygen diffusion is not a rate-limiting step in the SCC mechanism and that chromium diffusion in the bulk close the crack tip could be a key parameter. (authors)

  14. Recent bibliography on analytical and sampling problems of a PWR primary coolant Suppl. 3

    International Nuclear Information System (INIS)

    Illy, H.

    1985-03-01

    The present supplement to the bibliography on analytical and sampling problems of PWR primary coolant covers the literature published in 1984 and includes some references overlooked in the previous volumes dealing with the publications of the last 10 years. References are devided into topics characterized by the following headlines: boric acid; chloride; chlorine; carbon dioxide; general; gas analysis; hydrogen isotopes; iodine; iodide; nitrogen; noble gases and radium; ammonia; ammonium; oxygen; other elements; radiation monitoring; reactor safety; sampling; water chemistry. Under a given subject bibliographical information is listed in alphabetical order of the authors. (V.N.)

  15. Study of the formation and transport of corrosion products in PWR primary circuit simulators

    International Nuclear Information System (INIS)

    Noe, M.; Frejaville, G.; Camp, J.J.

    1983-01-01

    The formation, migration and deposition of corrosion products in PWR primary circuits are studied in out-of-reactor loops. The aim of these studies is to limit the build-up of the radiation fields impinging on out-of-flux walls and to reduce the danger of rapid corrosion of fuel cans, taking into account the tougher conditions imposed on current trends in the operation of such industrial plants. Four simulator loops and their respective possibilities and research methods are described. (author)

  16. Zinc Addition Effects on General Corrosion of Austenitic Stainless Steels in PWR Primary Conditions

    International Nuclear Information System (INIS)

    Qiao Peipeng; Zhang Lefu; Liu Ruiqin; Jiang Suqing; Zhu Fawen

    2010-01-01

    Zinc addition effects on general corrosion of austenitic stainless steel 316 and 304 were investigated in simulated PWR primary coolant without zinc or with 50 ppb zinc addition at 315 degree C for 500 h. The results show that with the addition of zinc, the corrosion rate of austenitic stainless steel is effectively reduced, the surface oxide film is thinner, the morphology and chemical composition of surface oxide scales are evidently different from those without zinc. There are needle-like corrosion products on the surface of stainless steel 304. (authors)

  17. The impact of radiolytic yield on the calculated ECP in PWR primary coolant circuits

    International Nuclear Information System (INIS)

    Urquidi-Macdonald, Mirna; Pitt, Jonathan; Macdonald, Digby D.

    2007-01-01

    A code, PWR-ECP, comprising chemistry, radiolysis, and mixed potential models has been developed to calculate radiolytic species concentrations and the corrosion potential of structural components at closely spaced points around the primary coolant circuits of pressurized water reactors (PWRs). The pH(T) of the coolant is calculated at each point of the primary-loop using a chemistry model for the B(OH) 3 + LiOH system. Although the chemistry/radiolysis/mixed potential code has the ability to calculate the transient reactor response, only the reactor steady state condition (normal operation) is discussed in this paper. The radiolysis model is a modified version of the code previously developed by Macdonald and coworkers to model the radiochemistry and corrosion properties of boiling water reactor primary coolant circuits. In the present work, the PWR-ECP code is used to explore the sensitivity of the calculated electrochemical corrosion potential (ECP) to the set of radiolytic yield data adopted; in this case, one set had been developed from ambient temperature experiments and another set reported elevated temperatures data. The calculations show that the calculated ECP is sensitive to the adopted values for the radiolytic yields

  18. SCC of cold-worked austenitic stainless steels exposed to PWR primary water conditions: susceptibility to initiation

    International Nuclear Information System (INIS)

    Herms, E.; Raquet, O.; Sejourne, L.; Vaillant, F.

    2009-01-01

    Heavily cold-worked austenitic stainless steels (AISI 304L and 316L types) could be significantly susceptible to Stress Corrosion Cracking (SCC) when exposed to PWR nominal primary water conditions even in absence of any pollutants. Susceptibility to SCC was shown to be related with some conditions such as initial hardness, procedure of cold-work or dynamic straining. A dedicated program devoted to better understand the initiation stage on CW austenitic stainless steels in PWR water is presented. Initiation is studied thanks to SCC test conditions leading to an intergranular cracking propagation mode on a CW austenitic stainless steel which is the mode generally reported after field experience. SCC tests are carried out in typical primary water conditions (composition 1000 ppm B and 2 ppm Li) and for temperature in the range 290 - 340 C. Material selected is 316L cold-worked essentially by rolling (reduction in thickness of 40%). Initiation tests are carried out under various stress levels with the aim to investigate the evolution of the initiation period versus the value of applied stress. SCC tests are performed on cylindrical notched specimens in order to increase the applied stress and allow accelerated testing without modify the exposure conditions to strictly nominal hydrogenated PWR water. Respective influences of cyclic/dynamic conditions on SCC initiation are presented and discussed. Dedicated interrupted tests help to investigate the behaviour of the crack initiation process. These SCC tests have shown that crack initiation could be obtained after a very short time under dynamic loading conditions on heavily pre-strained austenitic stainless steels. Actual results show that the most limiting stage of the cracking process on CW 316L seems to be the transition from slow transgranular propagation of surface initiated cracks to intergranular fast propagation through the thickness of the sample. The duration of this stage during crack initiation tests is

  19. Thermodynamics and the transport of corrosion products in PWR primary circuits

    International Nuclear Information System (INIS)

    Turner, D.J.

    1992-01-01

    It is argued that practically useful models for the activation, transport and deposition of corrosion products in PWR primary circuits can only be produced on the basis of an improved understanding of the chemical processes which control them. In particular, if a model is to make reliable predictions it is essential that its thermodynamic basis be sound. This is not the case with most current models which employ the erroneous concept of a corrosion product 'solubility'. In addition to the misuse of this term, other complications are discussed. These include the need to take account of the consequences of Gibbs' phase rule and the fact that, for mixed spinels, neither the concept of a thermodynamic solubility nor of a solubility product is valid. There is no reason to believe that measured apparent solubilities of nickel ferrites or spinel mixtures containing cobalt can give any direct guidance on the direction of transport of Ni or Co in PWR primary circuits. This is more likely to be determined by the distribution of stable and unstable ferrites and chromites than by any temperature coefficient of apparent solubility. Most of the transport of Ni and Co into and out of the core probably occurs as a consequence of either chemical or mechanical transients. Most important is likely to be the oxidative destruction and subsequent re-precipitation of chromites which occurs as a consequence of the oxygenated conditions employed during plant shutdown. (author)

  20. Stress corrosion cracking of Alloy 600 in primary water of PWR: study of chromium diffusion

    International Nuclear Information System (INIS)

    Chetroiu, Bogdan-Adrian

    2015-01-01

    Alloy 600 (Ni-15%Cr-10%Fe) is known to be susceptible to Stress Corrosion Cracking (SCC) in primary water of Pressurized Water Reactors (PWR). Recent studies have shown that chromium diffusion is a controlling rate step in the comprehension of SCC mechanism. In order to improve the understanding and the modelling of SCC of Alloy 600 in PWR primary medium the aim of this study was to collect data on kinetics diffusion of chromium. Volume and grain boundary diffusion of chromium in pure nickel and Alloy 600 (mono and poly-crystals) has been measured in the temperature range 678 K to 1060 K by using Secondary Ions Mass Spectroscopy (SIMS) and Glow Discharge-Optical Spectrometry (GD-OES) techniques. A particular emphasis has been dedicated to the influence of plastic deformation on chromium diffusion in nickel single crystals (orientated <101>) for different metallurgical states. The experimental tests were carried out in order to compare the chromium diffusion coefficients in free lattice (not deformed), in pre-hardening specimens (4% and 20%) and in dynamic deformed tensile specimens at 773 K. It has been found that chromium diffusivity measured in dynamic plastic deformed creep specimens were six orders of magnitude greater than those obtained in not deformed or pre-hardening specimens. The enhancement of chromium diffusivity can be attributed to the presence of moving dislocations generated during plastic deformation. (author)

  1. Verification test for radiation reduction effect and material integrity on PWR primary system by zinc injection

    Energy Technology Data Exchange (ETDEWEB)

    Kawakami, H.; Nagata, T.; Yamada, M. [Nuclear Power Engineering Corp. (Japan); Kasahara, K.; Tsuruta, T.; Nishimura, T. [Mitsubishi Heavy Industries, Ltd. (Japan); Ishigure, K. [Saitama Inst. of Tech. (Japan)

    2002-07-01

    Zinc injection is known to be an effective method for the reduction of radiation source in the primary water system of a PWR. There is a need to verify the effect of Zn injection operation on radiation source reduction and materials integrity of PWR primary circuit. In order to confirm the effectiveness of Zn injection, verification test as a national program sponsored by Ministry of Economy, Trade and Industry (METI) was started in 1995 for 7-year program, and will be finished by the end of March in 2002. This program consists of irradiation test and material integrity test. Irradiation test as an In-Pile-Test managed by AEAT Plc(UK) was performed using the LVR-15 reactor of NRI Rez in Check Republic. Furthermore, Out-of-Pile-Test using film adding unit was also performed to obtain supplemental data for In-Pile-Test at Takasago Engineering Laboratory of NUPEC. Material Integrity test was planned to perform constant load test, constant strain test and corrosion test at the same time using large scale Loop and slow strain extension rate testing (SSRT) at Takasago Engineering Laboratory of NUPEC. In this paper, the results of the verification test for Zinc program at present are discussed. (authors)

  2. Application of surface science to the study of the corrosion of PWR primary circuit materials

    International Nuclear Information System (INIS)

    Harris, S.J.

    1989-04-01

    This thesis describes a study of the corrosion and oxidation of PWR primary circuit materials using surface sensitive spectroscopic techniques. An X-ray photoemission spectroscopy (XPS) study of a number of mixed oxides of known composition is described and the information obtained is related to XPS measurements made on the surface of iron and nickel based alloys oxidised under controlled conditions. A secondary ion mass spectroscopy (SIMA) study on these mixed transition metal oxides is also described. The gaseous oxidation of stainless steel 3041 and Inconel-690 is examined. Both alloys were oxidised at 600K in air with the composition of the oxide films formed studied by a range of surface spectroscopic methods. Further experimental work was performed on Inconel-690 to examine the effects of surface pretreatment and the effects of low oxygen partial pressures on the formation of oxide films at 600 K. The incorporation of the radionuclide, cobalt-60, into the oxide films formed on structural components of a PWR, result in the build up of radiation fields. A method of pretreating the surface of the alloy stainless steel 3041, in order to reduce the level of cobalt adsorbed into the oxide film formed under simulated primary coolant conditions is examined and contrasts with treatments which have been developed to release cobalt adsorbed in existing oxide layers under reactor conditions are discussed. (author)

  3. Laboratory results gained from cold worked type 316Ti under simulated PWR primary environment

    International Nuclear Information System (INIS)

    Devrient, B.; Kilian, R.; Koenig, G.; Widera, M.; Wermelinger, T.

    2015-01-01

    Beginning in 2005, intergranular stress corrosion cracking (IGSCC) of barrel bolts made from cold worked type 316Ti (German Material No. 1.4571 K) was observed in several S/KWU type PWRs. This mechanism was so far less understood for PWR primary conditions. Therefore an extended joint research program was launched by AREVA GmbH and VGB e.V. to clarify the specific conditions which contributed to the observed findings on barrel bolts. In the frame of this research program beneath the evaluation of the operational experience also laboratory tests on the general cracking behavior of cold worked type 316Ti material, which followed the same production line as for barrel bolt manufacturing in the eighties, with different cold work levels covering up to 30 % were performed to determine whether there is a specific susceptibility of cold worked austenitic stainless steel specimens to suffer IGSCC under simulated PWR primary conditions. All these slow strain rate tests on tapered specimens and component specimens came to the results that first, much higher cold work levels than used for the existing barrel bolts are needed for IGSCC initiation. Secondly, additional high active plastic deformation is needed to generate and propagate intergranular cracking. And thirdly, all specimens finally showed ductile fracture at the applied strain rates. (authors)

  4. Application of liquid chromatography techniques to the measurement of soluble transition metals in PWR primary coolant

    International Nuclear Information System (INIS)

    Amey, M.D.H.; Brown, G.R.

    1987-01-01

    Two chromatographic techniques have been developed, and evaluated for the on-line analysis of soluble transition metals, particularly cobalt, in PWR primary coolant. Automatic operation and control, together with data processing and storage has been achieved by interfacing a Dionex ion chromatograph to a microprocessor control system. An absolute detection limit of 0.1 ng cobalt has been obtained which, with on-line sample preconcentration (100 ml), has enabled measurements to be made down to part-per-trillion levels (0.001 ppb). Application of the techniques to PWR coolant analysis was demonstrated by a programme of work on the Half Megawatt Loop at Winfrith. During this work some aspects of the behaviour of soluble metal species have been studied in both de-oxygenated and hydrogenated conditions. The effects of changes in coolant chemistry, operating temperature, and sample line flowrates on circulating impurity levels are reported, together with the dramatic effects observed when part of the circuit pipework was replaced with new stainless steel tubing. (author)

  5. On the failure probability of the primary piping of the PWR

    International Nuclear Information System (INIS)

    Schueller, G.I.; Hampl, N.C.

    1984-01-01

    A methodology for quantification of the structural reliability of the primary piping (PP) of a PWR under operational and accidental conditions is developed. Biblis B is utilized as reference plant. The PP structure is modeled utilizing finite element procedures. Based on the properties of the operational and internal accidental conditions, a static analysis suffices. However, a dynamic analysis considering non-linear effects of the soil-structure-interaction is to be used to determine load effects due to earthquake induced loading. Considering realistically the presence of initial cracks in welds and considering annual frequencies of occurrence of the various loading conditions, a crack propagation calculation utilizing the Forman model is carried out. Simultaneously leak and break probabilities using the 'Two Criteria'-Aproach are computed. A Monte Carlo simulation procedure is used as method of solution. (Author) [pt

  6. Stress corrosion cracking of steam generator tube and primary pipe in PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Weiguo; Gao Fengqin; Zhou Hongyi

    1992-03-01

    The behavior of stress corrosion cracking (SCC) was studied by slow strain rate test (SSRT), constant load test (CLT) and low frequency cyclic loading test (LFCLT). The purpose of these tests is to get the test data for evaluating the integrity of pressurized boundary of pipes in Qinshan and Guangdong nuclear power plants (NPPs). Tested materials are 316 nuclear grade stainless steel (SS) for primary pipes in welded heat affected zone (WHAZ) and tubes of heat transfer, such as Incoloy-800, Inconel-600 and 321 SS which are used for steam generator in PWR NPPs. The effects of material metallurgy, shot peening treatment, tensile load, strain rate, cyclic load and water chemistry on the behavior of SCC were considered

  7. Stress corrosion cracking of steam generator tube and primary pipe in PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Weiguo; Gao Fengqin; Zhou Hongyi

    1993-01-01

    The behavior of stress corrosion cracking (SCC) is studied by slow strain rate test (SSRT), constant load test (CLT) and low frequency cyclic loading test (LFCLT). The purpose of these tests is to get the test data for evaluating the integrity of pressurized boundary of pipes in Qinshan and Guangdong nuclear power plants. Tested materials are 316 nuclear grade stainless steel (SS) for primary pipes in welded heat affected zone (WHAZ) and steam generator tubes, such as Incoloy-800, Inconel-600, Inconel-690 and 321 SS which are used for steam generator in PWR. The effects of material metallurgy, shot-peening treatment, tensile load, strain rate, cyclic load and water chemistry on the behavior of SCC are investigated

  8. Influence of deformation on SCC susceptibility of austenitic stainless steel in PWR primary water

    Energy Technology Data Exchange (ETDEWEB)

    Kaneshima, Yoshiari; Totsuka, Nobuo; Nakajima, Nobuo [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2001-09-01

    Slow strain rate tests (SSRT) were carried out to evaluate the SCC susceptibility of four types of austenitic stainless steels (SUS304, SUS316, SUS304L and SUS316L) in PWR primary water. The influence of deformation on SCC susceptibility of SUS316 was studied. All types of stainless steel were susceptible to SCC, and the SCC susceptibility varied depending on the steel type. The comparison of the SSRT results and tensile test in air based on the reduction of area measurement showed that the SCC susceptibility increased with increasing the degree of deformation. For explaining the influence of deformation on SCC susceptibility, it is necessary to evaluate both intergranular and transgranular fractures. (author)

  9. An improvement of estimation method of source term to the environment for interfacing system LOCA for typical PWR using MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Han, Seok Jung; Kim, Tae Woon; Ahn, Kwang Il [Risk and Environmental Safety Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-06-15

    Interfacing-system loss-of-coolant-accident (ISLOCA) has been identified as the most hazardous accident scenario in the typical PWR plants. The present study as an effort to improve the knowledge of the source term to the environment during ISLOCA focuses on an improvement of the estimation method. The improvement was performed to take into account an effect of broken pipeline and auxiliary building structures relevant to ISLOCA. An estimation of the source term to the environment was for the OPR-1000 plants by MELOCR code version 1.8.6. The key features of the source term showed that the massive amount of fission products departed from the beginning of core degradation to the vessel breach. The release amount of fission products may be affected by the broken pipeline and the auxiliary building structure associated with release pathway.

  10. Major activated corrosion products cobalt, silver and antimony in the primary coolant of PWR power plants

    International Nuclear Information System (INIS)

    Xu Mingxia

    2012-01-01

    The production of the major activated corrosion products such as cobalt, silver and antimony in the primary coolant of PWR power plants and the impacts on the increase of the dose rates caused by these corrosion products during the shutdown are described in the paper. Investigating the corrosion product behavior during the operation and shutdown periods aims at detecting the appearance of these radiological pollutants in the early time and searching relevant solutions that may enable eventually to decrease the dose rate. The solutions may include: Replacing critical material in the primary system's equipment and components, which contact with primary coolant circuit to possibly limit the source term, Elaborating strictly the specific chemical and shutdown procedure to optimize the purification capacity and to minimize the over-contaminations; Improving purification techniques according to the real operation circumstance, and limiting the impacts of these pollutants. It is obvious in the real practices that implementing appropriate solution will be benefit to decrease or limit the pollutants species like cobalt, silver and antimony. (author)

  11. Low cycle fatigue of Alloy 690 and welds in a simulated PWR primary water environment

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jongdae; Cho, Pyungyeon; Jang, Changheui [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Cho, Pyungyeon [Khalifa Univ., Abu Dhabi (United Arab Emirates); Kim, Tae Soon; Lee, Yong Sung [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2013-05-15

    In this study, environmental fatigue tests for these materials were performed and the new prediction model of fatigue life of Alloy 690 and weld in primary water condition was proposed. To evaluate the fatigue life of Alloy 690 and 52M in a PWR environment, low cycle fatigue tests were performed and revised fatigue life prediction models and environmental factor were proposed. With the revised Fen model for Alloy 690 and 52M, the reliability of the fatigue life prediction has been improved. The reduction of low cycle fatigue life of metallic materials in the primary coolant water environments has been the subject of debate between the utility and regulator since 1980s. It became the significant licensing problem since the issue of RG-1.207 by U. S. NRC. The statistical model for the environmental factor, Fen, specified in RG-1.207 was based on the extensive test results accumulated by the ANL and Japanese national program. Of the materials, the limited fatigue life data of Ni-Cr-Fe alloys were used to develop the Fen for the alloys. Furthermore, test data for Alloy 690 and its weld are limited. Considering that Alloy 690 will be extensively used in the new nuclear power plants, additional effort to validate or improve current Fen model is required.

  12. VGB primary and secondary side water chemistry guidelines for PWR plants

    International Nuclear Information System (INIS)

    Neder, H.; Wolter, D.; Staudt, U.

    2007-01-01

    The recent revision of the VGB Water Chemistry Guidelines was issued in 2005 and published in the second half of 2006. These guidelines are based on the primary and secondary side operating chemistry experience with all Siemens designed pressurized water reactors gained since the beginning of the 1980s. These guidelines cover For the primary side chemistry Modified lithium boron chemistry, Zinc chemistry for dose rate reduction, Enriched boric acid (EBA) chemistry for high duty core design For the secondary side chemistry High all-volatile treatment (AVT) chemistry (high pH operation) Oxygen injection in the secondary side Especially for the secondary side chemistry, compared with the water chemistry guidelines of other organizations worldwide, these Guidelines are less stringent, providing more operational flexibility to the plant operation, and can be applied for all new designs of steam generators with egg-crates or broached hole tube supports and with I 690TT or I 800 tubing materials. This paper gives an overview of the 2006 revision of the VGB Water Chemistry Guidelines for PWR plants and describes the fundamental goals of water chemistry operation strategies. In addition, the reasons for the selected control parameters and action levels, to achieve an adequate plant performance, are presented based on the operating experience. (orig.)

  13. Modeling in fast dynamics of accidents in the primary circuit of PWR type reactors

    International Nuclear Information System (INIS)

    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)

  14. SCC growth behaviors of austenitic stainless steels in simulated PWR primary water

    Science.gov (United States)

    Terachi, T.; Yamada, T.; Miyamoto, T.; Arioka, K.

    2012-07-01

    The rates of SCC growth were measured under simulated PWR primary water conditions (500 ppm B + 2 ppm Li + 30 cm3/kg-H2O-STP DH2) using cold worked 316SS and 304SS. The direct current potential drop method was applied to measure the crack growth rates for 53 specimens. Dependence of the major engineering factors, such as yield strength, temperature and stress intensity was systematically examined. The rates of crack growth were proportional to the 2.9 power of yield strength, and directly proportional to the apparent yield strength. The estimated apparent activation energy was 84 kJ/mol. No significant differences in the SCC growth rates and behaviors were identified between 316SS and 304SS. Based on the measured results, an empirical equation for crack growth rate was proposed for engineering applications. Although there were deviations, 92.8% of the measured crack growth rates did not exceed twice the value calculated by the empirical equation.

  15. 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.

  16. Influence of Localized Plasticity on IASCC Sensitivity of Austenitic Stainless Steels under PWR Primary Water

    Science.gov (United States)

    Cissé, Sarata; Tanguy, Benoit; Laffont, Lydia; Lafont, Marie-Christine; Guerre, Catherine; Andrieu, Eric

    The sensibility of precipitation-strengthened A286 austenitic stainless steel to Stress Corrosion Cracking (SCC) is studied by means of Slow Strain Rate Tests (SSRT). First, alloy cold working by Low Cycle Fatigue (LCF) is investigated. Fatigue tests under plastic strain control are performed at different strain levels (Δ ɛp/2=0.2%, 0.5% and 0.8%) in order to establish correlation between stress softening and deformation microstructure resulting from LCF tests. Deformed microstructures have been identified through TEM investigations. Three states of cyclic behaviour for precipitation-strengthened A286 have been identified: hardening, cyclic softening and finally saturation of softening. It is shown that the A286 alloy cyclic softening is due to microstructural features such as defects — free deformation bands resulting from dislocations motion along family plans , that swept defects or γ' precipitates and lead to deformation localization. In order to quantify effects of plastic localized deformation on intergranular stress corrosion cracking (IGSCC) of the A286 alloy in PWR primary water, slow strain rate tests are conducted. For each cycling conditions, two specimens at a similar stress level are tested: the first containing free precipitate deformation bands, the other not significant of a localized deformation state. SSRT tests are still in progress.

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

    International Nuclear Information System (INIS)

    Han, Jeong Ho; Nyo, Kye Ho; Lee, Deok Hyun; Lim, Deok Jae; Ahn, Jin Keun; Kim, Sun Jin

    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)

  18. Structural analysis of surface film on alloy 600 formed under environment of PWR primary water

    Energy Technology Data Exchange (ETDEWEB)

    Terachi, Takumi; Totsuka, Nobuo; Yamada, Takuyo; Nakagawa, Tomokazu [Inst. of Nuclear Safety System Inc., Mihama, Fukui (Japan); Deguchi, Hiroshi [Kansai Electric Power Co., Inc., Osaka (Japan); Horiuchi, Masaki; Oshitani, Masato [Kanden Kako Co., Ltd., Osaka (Japan)

    2002-09-01

    It has been shown by one of the present authors and so forth that PWSCC of alloy 600 relates to dissolved hydrogen concentration (DH) in water and oxide film structure. However, the mechanism of PWSCC has not been clear yet. Therefore, in order to investigate relationship between them, structural analysis of the oxide film formed under the environment of PWR primary water was carried out by using X-ray diffraction, the scanning electron microscope and the transmission electron microscope. Especially, to perform accurate analysis, the synchrotron orbital radiation with SPring-8 was tried to use for thin film X-ray diffraction measurement. From the results, observed are as follows: 1. the oxide film is mainly composed of NiO, under the condition without hydrogen. 2. In the environment of DH 2.75ppm, the oxide film forms thin spinel structures. 3. On the other hand, needlelike oxides are formed at DH 1ppm. For this reason, around 1ppm of DH there would be the boundary that stable NiO and spinel oxide generate, and it agrees with the peak range of the PWSCC susceptibility on hydrogen. From this, it is suggested that the boundary of NiO/spinel oxide affects the SCC susceptibility. (author)

  19. Influence of dissolved hydrogen on oxide film and PWSCC of Alloy 600 in PWR primary water

    Energy Technology Data Exchange (ETDEWEB)

    Nakagawa, Tomokazu; Totsuka, Nobuo; Nakajima, Nobuo [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2001-09-01

    In order to investigate the influence of dissolved hydrogen (DH) on the corrosion behavior and PWSCC of Alloy 600 in primary water of PWR under actual operating temperature range, we carried out electrochemical polarization measurement, repassivation test, analysis of the oxide film on the alloy by AES, XPS and PWSCC test. In all cases, the content of DH was changed from 0 to 45 cc/kgH{sub 2}O. The anodic polarization curve reveals that the peak current density increases with increasing DH. The result of the repassivation test shows that the repassivation rate decreases with increasing DH, and the changes of the above two become larger between 11 and 22 cc/kgH{sub 2}O of DH. According to the results of oxide film analysis, it is seen that the oxide films formed below 11 cc/kgH{sub 2}O of DH are relatively thick and rich in Ni, but those formed at higher DH contents are relatively thin and rich in Cr and Fe. The susceptibility of the alloy to PWSCC has a peak at 11 cc/kgH{sub 2}O of DH, which reveals that the property of the oxide film may play important role in PWSCC of alloy. (author)

  20. Analysis of corrosion product transport in PWR primary system under non-convective condition

    International Nuclear Information System (INIS)

    Han, Byoung Sub

    1992-02-01

    Product TRANsport), which can predict the corrosion product and radioactivity transport within the primary coolant system, and also can be utilized for the computer simulation with actual plant data of currently operating Korean nuclear power plants to predict the transport of the radionuclides. In this study, the following problems will be updated, improved and compared with the already existing codes: 1) development and analysis of recent mechanistic modelling of corrosion product deposition, 2) application and modification due to the temperature kinetic effect, 3) separation of the effect of Fe, Co, Ni and Mn solubility rather than Fe solubility alone, and 4) consideration of Ni activation and recoil process. By applying the above updated and improved mechanisms, the corrosion product behavior in PWR of currently operating Korean unclear power plants has been simulated. In addition, the evaluation of particulate transport, independent solubility data of major radionuclides and acute nodalization were included and extended. Then, with the developed computer code, we have evaluated and analyzed the activity and corrosion product build-up controlled by many parameters such as pH, composition of metal, and auxiliary system performance

  1. BWR water chemistry guidelines and PWR primary water chemistry guidelines in Japan – Purpose and technical background

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, Hirotaka, E-mail: kawamuh@criepi.denken.or.jp [Central Research Institute of Electric Power Industry (Japan); Hirano, Hideo [Central Research Institute of Electric Power Industry (Japan); Katsumura, Yousuke [University of Tokyo (Japan); Uchida, Shunsuke [Tohoku University (Japan); Mizuno, Takayuki [Mie University (Japan); Kitajima, Hideaki; Tsuzuki, Yasuo [Japan Nuclear Safety Institute (Japan); Terachi, Takumi [Institute of Nuclear Safety System, Inc. (Japan); Nagase, Makoto; Usui, Naoshi [Hitachi-GE Nuclear Energy, Ltd. (Japan); Takagi, Junichi; Urata, Hidehiro [Toshiba Corporation (Japan); Shoda, Yasuhiko; Nishimura, Takao [Mitsubishi Heavy Industry, Ltd. (Japan)

    2016-12-01

    Highlights: • Framework of BWR/PWR water chemistry Guidelines in Japan are presented. • Guideline necessity, definitions, philosophy and technical background are mentioned. • Some guideline settings for control parameters and recommendations are explaines. • Chemistry strategy is also mentioned. - Abstract: After 40 years of light water reactor (LWR) operations in Japan, the sustainable development of water chemistry technologies has aimed to ensure the highest coolant system component integrity and fuel reliability performance for maintaining LWRs in the world; additionally, it aimed to achieve an excellent dose rate reduction. Although reasonable control and diagnostic parameters are utilized by each boiling water reactor (BWR) and pressurized water reactor (PWR) owner, it is recognized that specific values are not shared among everyone involved. To ensure the reliability of BWR and PWR operation and maintenance, relevant members of the Atomic Energy Society of Japan (AESJ) decided to establish guidelines for water chemistry. The Japanese BWR and PWR water chemistry guidelines provide strategies to improve material and fuel reliability performance as well as to reduce dosing rates. The guidelines also provide reasonable “control values”, “diagnostic values” and “action levels” for multiple parameters, and they stipulate responses when these levels are exceeded. Specifically, “conditioning parameters” are adopted in the Japanese PWR primary water chemistry guidelines. Good practices for operational conditions are also discussed with reference to long-term experience. This paper presents the purpose, technical background and framework of the preliminary water chemistry guidelines for Japanese BWRs and PWRs. It is expected that the guidelines will be helpful as an introduction to achieve safety and reliability during operations.

  2. Assessment of cracked pipes in primary piping systems of PWR nuclear reactors

    International Nuclear Information System (INIS)

    Jong, Rudolf Peter de

    2004-01-01

    Pipes related to the Primary System of Pressurized Water Reactors (PWR) are manufactured from high toughness austenitic and low alloy ferritic steels, which are resistant to the unstable growth of defects. A crack in a piping system should cause a leakage in a considerable rate allowing its identification, before its growth could cause a catastrophic rupture of the piping. This is the LBB (Leak Before Break) concept. An essential step in applying the LBB concept consists in the analysis of the stability of a postulated through wall crack in a specific piping system. The methods for the assessment of flawed components fabricated from ductile materials require the use of Elasto-Plastic Fracture Mechanics (EPFM). Considering that the use of numerical methods to apply the concepts of EPFM may be expensive and time consuming, the existence of the so called simplified methods for the assessment of flaws in piping are still considered of great relevance. In this work, some of the simplified methods, normalized procedures and criteria for the assessment of the ductile behavior of flawed components available in literature are described and evaluated. Aspects related to the selection of the material properties necessary for the application of these methods are also discussed. In a next .step, the methods are applied to determine the instability load in some piping configurations under bending and containing circumferential through wall cracks. Geometry and material variations are considered. The instability loads, obtained for these piping as the result of the application of the selected methods, are analyzed and compared among them and with some experimental results obtained from literature. The predictions done with the methods demonstrated that they provide consistent results, with good level of accuracy with regard to the determination of maximum loads. These methods are also applied to a specific Study Case. The obtained results are then analyzed in order to give

  3. Magnetite solubility studies under simulated PWR primary-side conditions, using lithiated, hydrogenated water

    International Nuclear Information System (INIS)

    Hewett, John; Morrison, Jonathan; Cooper, Christopher; Ponton, Clive; Connolly, Brian; Dickinson, Shirley; Henshaw, Jim

    2014-01-01

    As software for modelling dissolution, precipitation, and transport of metallic species and subsequent CRUD deposition within nuclear plant becomes more advanced, there is an increasing need for accurate and reliable thermodynamic data. The solubility behaviour of magnetite is an example of such data, and is central to any treatment of CRUD solubility due to the prevalence of magnetite and nickel ferrites in CRUD. Several workers have shown the most consistent solubility data comes from once-through flowing systems. However, despite a strong consensus between the results in acidic to mildly alkaline solutions, there is disagreement between the results at approximately pH 25C 9 and higher. A programme of experimental work is on-going at the University of Birmingham, focusing on solubility of metal oxides (e.g., magnetite) in conditions relevant to PWR primary coolant. One objective of this programme is to calculate thermodynamic constants from the data obtained. Magnetite solubility from 200 to 300°C, in lithiated, hydrogenated water of pH 25C 9–11 is being studied using a once-through rig constructed of 316L stainless steel. The feedwater is pumped at 100 bar pressure through a heated bed of magnetite granules, and the output solution is collected and analysed for iron and several other metals by ICP-MS. This paper presents results from preliminary tests without magnetite granules, in which the corroding surface of the rig itself was used as the sole source of soluble iron and of dissolved hydrogen. Levels of iron were generally within an order of magnitude of literature solubility values. Comparison of results at different flow rates and temperatures, in conjunction with conclusions drawn from the published literature, suggests that this is likely due to the presence of particulate matter in a greatly under-saturated solution, compensating for the low surface area of oxide in contact with the solution. (author)

  4. Fatigue crack growth threshold of austenitic stainless steels in simulated PWR primary water

    International Nuclear Information System (INIS)

    Tsutsumi, Kazuya; Yamamoto, Kenji; Nitta, Yoshikazu

    2007-01-01

    Many studies have revealed that fatigue crack growth (FCG) rate of austenitic stainless steels is accelerated in light water reactor environment compared to that in air at room temperature. Major driving factors in the acceleration of FCG rate are stress ratio, temperature and stress rise time. Based on this knowledge, FCG curves have been developed considering these factors as parameters. However, there are few data of FCG threshold ΔK th in light water reactor environment. Hence it is necessary to clarify FCG rate under near-threshold condition for more accurate evaluation of fatigue crack growth behavior under cyclic stress with relatively low ΔK. In the present study, therefore, ΔK th was determined for austenitic stainless steels in simulated PWR primary water, and FCG behavior under near-threshold condition was revealed by collecting fatigue crack propagation data. The results are summarized as follows: No propagation of fatigue crack was found in high temperature water, and there was a definite ΔK th . Average ΔK eff,th was 4.3 MPa·m 0.5 at 325degC, 3.3 MPa·m 0.5 at 100degC, and there was no considerable reduction compared to currently known ΔK eff,th in air. Thus, it was revealed tha ambient conditions had minimal effect, on ΔK eff,th , ΔK th increases with increasing temperature and decreasing frequency. As a result of fracture surface observation, oxide-induced-crack-closure was considered to be a cause of the dependency described above. In addition, it was suggested that changes in material properties also had influence on ΔK th, since ΔK eff,th itself increased at elevated temperature. (author)

  5. Environmental effect on cracking of an 304L austenitic stainless steels in PWR primary environment under cyclic loading

    International Nuclear Information System (INIS)

    Huin, N.

    2013-01-01

    The present study was undertaken in order to get further insights on cracking mechanisms in a 304L stainless steel. More precisely, a first objective of this study was to evaluate the effect of various cold working conditions on the cyclic stress-strain behavior and the fatigue life in air and in PWR primary environment. In air a prior hardening was found to reduce the fatigue life in the LCF regime but not in primary environment. In both environments, the fatigue limit of the hardened materials was increased after cold working.The second objective addresses the effect of the air and the PWR primary environments on the cracking mechanisms (initiation and propagation) in the annealed material in the LCF regime. More precisely, the kinetics of crack initiation and micro crack propagation were evaluated with a multi scale microscopic approach in air and in primary environment. In PWR primary environment, during the first cycles, preferential oxidation occurs along emerging dissociated dislocation and each cycle generates a new C-rich/Fe-rich oxide layer. Then, during cycling, the microstructure evolves from stacking fault into micro twinning and preferential oxidation occurs by continuous shearing and dissolution of the passive film. Beyond a certain crack depth (≤3 μm), the crack starts to propagate with a direction close to a 90 degrees angle from the surface. The crack continues its propagation by successive generation of shear bands and fatigue striations at each cycle up to failure. The role of corrosion hydrogen on these processes is finally discussed. (author)

  6. Appropriate zinc addition management into PWR primary coolant after the plant long-term maintenance

    International Nuclear Information System (INIS)

    Hirose, Atsushi; Matsui, Ryo; Imamura, Haruki; Takahashi, Akira; Shimizu, Yuichi; Kogawa, Noritaka; Nagamine, Kunitaka

    2014-01-01

    Zinc addition into the PWR primary coolant is known as an effective method to reduce the radioactivity build up. The reduction effect has been confirmed by actual plant experience of the Genkai Nuclear Power Plant Unit 1 to 4 and the Sendai Nuclear Power Plant Unit 1 to 2 which are operated by Kyushu Electric Power Co. in Japan. Zinc addition is suspended at shut-down, and is resumed after heat up or arrival at full power. In usual maintenance, the period when zinc addition is not applied is short; thus it is considered that suspension of zinc addition does not have practical influence on the corrosion and the radioactivity buildup in the oxide layer of surface for the primary equipment and piping. On the other hand, in case the maintenance period is much longer, the new oxide which does not contain zinc has grown, and then the structure of the oxide layer may be changed. Therefore, it is considered that zinc addition suspension in long-term period has possibilities to deteriorate the dose reduction effect. In order to verify the effect of long-term suspension of zinc addition upon oxide layer, the lab experiment was carried out using TT690 alloy which is the constitution material of the steam generator tubes under the conditions of long-term and the subsequent resuming operations. After the experiment, the specimens were analyzed by IMA and chemical analysis. These measurement results suggest the difference of the oxide layer is little or none between long-term suspension of zinc addition and short-term suspension of zinc suspension. Hence it is considered that influence of long-term maintenance on the oxide layer is small. Furthermore, in this study, in order to evaluate the influence of the suspension of zinc addition in the operation period, specimens of oxide film formed with zinc were carried out the corrosion test in the simulated RCS condition without zinc. These measurement results indicate the effect of reduction of the activity build up will become less

  7. Development status of nuclear power in China and fundamental research progress on PWR primary water chemistry in China

    International Nuclear Information System (INIS)

    Wu, Xinqiang; Liu, Xiahe; Han, En-Hou; Ke, Wei; Xu, Yuming

    2015-01-01

    China's non-fossil fuels are expected to reach 20% in primary energy ratio by 2030. It is urgent for China to speed up the development of nuclear power to increase energy supply, reduce gas emissions and optimize resource allocation. Chinese government slowed down the approval of new nuclear power plant (NPP) projects after Fukushima accident in 2011. At the end of 2012, the State Council approved the nuclear safety program and adjusted long-term nuclear power development plan (2011-2020), the new NPP's projects have been restarted. In June 2015, there are 23 operating units in mainland in China with total installed capacity of about 21.386 GWe; another 26 units are under construction with total installed capacity of 28.5 GWe. The main type of reactors in operation and under construction in China is pressurized water reactor (PWR), including the first AP1000 NPPs in the world (units 1 in Sanmen) and China self-developed Hualong one NPPs (units 5 and 6 in Fuqing). Currently, China's nuclear power development is facing historic opportunities and also a series of challenges. One of the most important is the safety and economy of nuclear power. The optimization of primary water chemistry is one of the most effective ways to minimize radiation field, mitigate material degradation and maintain fuel performance in PWR NPPs, which is also a preferred path to achieve both safety and economy for operating NPPs. In recent years, an increased attention has been paid to fundamental research and engineering application of PWR primary water chemistry in China. The present talk mainly consists of four parts: (1) development status of China's nuclear power industry; (2) safety of nuclear power and operating water chemistry; (3) fundamental research progress on Zn-injected water chemistry in China; (4) summary and future. (author)

  8. Behaviour of radiation fields in the Spanish PWR by the changes in coolant chemistry and primary system materials

    International Nuclear Information System (INIS)

    Llovet, R.; Fernandez Lillo, E.

    1995-01-01

    The Spanish PWR Owners Group established a program to evaluate the behavior of ex-core radiation fields and discriminate the effects of changes in coolant chemistry and primary system materials. Data from Vandellos, Asco, Almaraz and Trillo NPPs were analyzed Vandellos 2 was chosen as the lead plant and its data were thoroughly studied. The dose-rates evolution could be explained at each plant as a consequence of this sucessful program.Actions derived from the developed knowledge on this field have produced the stabilization or even reduction of radiation fields at these plants

  9. Probabilistic study of primary pump trip in a P.W.R. reactor: use of response surface methodology

    International Nuclear Information System (INIS)

    Bars, C.; Duchemin, B.; Maigret, N.; Peltier, J.; Rostan, O.; Villeneuve, M.J. de; Lanore, J.M.

    1981-09-01

    This paper describes a probabilistic study about the consequences of the trip or blockage of one of the three PWR reactor primary pumps. The distribution of the input parameters is taken into account and the resulting distribution of the consequence (number of failed fuel rods) is assessed. The necessity to do this study with the response surface methodology and the precautions to take are outlined. The results show that the probability to have failed fuel rods is about 10 -4 for pump trip and 0.16 for blockage with, in this case, a mean of 196 failed rods, that is 0.5 % of total number of rods

  10. Primary circuit leak detection an application on PWR vessel head penetrations

    International Nuclear Information System (INIS)

    Loisy, F.; Germain, J.L.; Chauvel, L.

    1996-01-01

    In 1991, cracks were discovered and localized in the lower part of certain vessel head adapters in EDF PWR units. While awaiting the replacement of the vessel heads in question, EDF developed systems to enable continuous monitoring of vessel head penetration, by means of early detection of leaks. One of these systems in based on detection of water vapour in a confined space above the vessel head. The efficiency of the measurement chain is particularly dependent on dilution of the leakage in the confined space prior TO entry in the sampling circuit. The detection threshold for this method is on the order of 1.2 liters/hour for a dilution rate of 1500 rate of 1500 m 3 /h and a dew point of 22 deg C. This system has now been in operation on three 1300-MW PWR units for three years, and has proved to function satisfactorily. (authors)

  11. Layout of the primary circuit with its components for PWR and BWR

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1981-01-01

    The light water-moderated and cooled pressurized water reactors and boiling water reactors constitute the basis of economic utilization of nuclear energy all over the world. Pressurized water reactors up to capacities of 3,800 MWth are those most used for power generation. However, their potential capacities exceed 3,800 MWth, so that already in the near future PWR are conseivable which readily generate 1,500 to 2,000 MWe. The main problem for starting the next generation of PWRs are of safety measure and licensing questions. Interesting applications of the PWRs are nuclear district heating, generation of process steam of desalination plants, steam injection into the ground for oil production or chemical factories. A new generation of natural circulation boiling water reactors with a capacity of 200 to 400 MW will be used for development of small industrial areas or for countries without an integral grid system. The natural circulation boiling water reactor will be subject of a separate lecture. Due to the fact of the majority of the PWR all over the world this lecture will discuss mainly PWR design aspects. (orig./RW)

  12. Crack growth testing of cold worked stainless steel in a simulated PWR primary water environment to assess susceptibility to stress corrosion cracking

    International Nuclear Information System (INIS)

    Tice, D.R.; Stairmand, J.W.; Fairbrother, H.J.; Stock, A.

    2007-01-01

    Although austenitic stainless steels do not show a high degree of susceptibility to stress corrosion cracking (SCC) in PWR primary environments, there is limited evidence from laboratory testing that crack propagation may occur under some conditions for materials in a cold-worked condition. A test program is therefore underway to examine the factors influencing SCC propagation in good quality PWR primary coolant. Type 304 stainless steel was subjected to cold working by either rolling (at ambient or elevated temperature) or fatigue cycling, to produce a range of yield strengths. Compact tension specimens were fabricated from these materials and tested in simulated high temperature (250-300 o C) PWR primary coolant. It was observed that the degree of crack propagation was influenced by the degree of cold work, the crack growth orientation relative to the rolling direction and the method of working. (author)

  13. 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.

  14. Effect of hardening on the crack growth rate of austenitic stainless steels in primary PWR conditions

    International Nuclear Information System (INIS)

    Castano, M.L.; Garcia, M.S.; Diego, G. de; Gomez-Briceno, D.; Francia, L.

    2002-01-01

    Intergranular cracking of non-sensitized materials, found in light water reactor (LWR) components exposed to neutron radiation, has been attributed to Irradiation Assisted Stress Corrosion Cracking (IASCC). Cracking of baffle former bolts, fabricated of AISI-316L and AISI-347, have been reported in some Europeans and US PWR plants. Examinations of removed bolts indicate the intergranular cracking characteristics can be associated with IASCC phenomena. Neutron radiation produce critical modifications of the microstructure and microchemical of stainless steels such hardening due to irradiation and Radiation Induce Segregation (RIS) at grain boundaries, among others. Chromium depletion at grain boundary due to RIS seems to justify the intergranular cracking of irradiated materials, both in plant and in lab tests, at high electrochemical corrosion potential (BWR-NWC environments), but it is not enough to explain cracking at low corrosion potential (BWR-HWC and PWR environments). In these latter conditions, hardening is considered a possible additional mechanism to explain the behavior of irradiated material. Radiation Hardening can be simulated in non irradiated material by mechanical deformation. Although some differences exists in the types of defects produced by radiation and mechanical deformation, it is accepted that the study of the stress corrosion behavior of unirradiated austenitic steels with different hardening levels would contribute to the understanding of IASCC mechanism. In order to evaluate the influence of hardening on the stress corrosion susceptibility of austenitic steels, crack growth rate tests with 316L and 347 stainless steels with nominal yield strengths from 500 to 900 MPa, produced by cold work are being carried out at 340 deg C in PWR conditions. Preliminary results indicate that crack propagation was obtained in the 316Lss and 347ss cold worked, even with a yield strength of 550 MPa. (authors)

  15. The application of transition metal ion chromatography to the determination of elemental and radiochemical species in PWR primary coolant

    International Nuclear Information System (INIS)

    Bridle, D.A.; Brown, G.R.; Johnson, P.A.V.

    1992-01-01

    The accurate determination of both elemental and radiochemical transition metal corrosion products, particularly cobalt and nickel, in PWR coolants is necessary if the transport mechanisms and their role in the development of out-of-core radiation fields are to be fully understood. AEA Technology, Winfrith, has collaborated for several years with a number of PWR utilities in Europe, developing advanced sampling and analytical techniques for the determination of both soluble and insoluble corrosion products in primary coolant. The design and installation of continuously flowing isokinetic capillary modifications to the existing sampling systems has been shown to be an effective method of providing a low, but representative, sample flow from high pressure systems for on-line determination of corrosion product species. Transition metal ion chromatography coupled with gamma-spectrometry has been used to determine both insoluble and soluble elemental and radiochemical species in reactor coolant, with particular attention being given to the determination of soluble elemental cobalt at levels as low as 1 ng per kg. Soluble species were determined directly following their concentration from up to 1 litre of coolant. Insoluble species collected on 0.45 micron filter membranes, following filtration of up to 1500 litres of coolant, were solubilised by fusion with potassium hydrogen sulphate before the application of ion chromatography. In each case the eluant from the chromatographic column was collected and the radionuclides determined by gamma-spectrometry

  16. Relationship between stress corrosion cracking and low frequency fatigue-corrosion of alloy 600 in PWR primary water

    International Nuclear Information System (INIS)

    Bosch, C.

    1998-01-01

    Stress corrosion cracking of PWR vessel head adapters is a main problem for nuclear industry. With the aim to better understand the influence of the mechanical parameters on the cracking phenomena (by stress corrosion (SCC) or fatigue corrosion (FC)) of alloy 600 exposed to primary PWR coolant, a parametrical study has been carried out. Crack propagation tests on CT test specimens have been implemented under static loads (stress corrosion tests) or low frequency cyclic loads (fatigue corrosion tests). Results (frequency influence, type of cycles, ratio charge on velocities and propagation modes of cracks) have allowed to characterize the transition domain between the crack phenomena of SCC and FC. With the obtained results, it has been possible too to differentiate the effects due to environmental factors and the effects due to mechanical factors. At last, a quantitative fractographic study and the observations of the microstructure at the tip of crack have led to a better understanding of the transitions of the crack propagation mode between the SCC and the FC. (O.M.)

  17. Three dimensional calculations of the primary coolant flow in a 900 MW PWR vessel. Steady state and transients computations

    International Nuclear Information System (INIS)

    Martin, A.; Alvarez, D.; Cases, F.

    1996-03-01

    The paper explains the chronological account and the first results obtained in the R and D program on the mixing in the 900 MW PWR vessels. After the presentation of the plant type simulated, we define the numerical tool, the (Finite Element Modelling) FEM N3S code. Two results are presented with a comparison with the experiment results issued of the BORA BORA mock up. The first case is dealing with the isothermal steady state mixing in the vessel with the three loops mass flow rate balanced. This case identified as a validation of our numerical tool shows a good agreement. The second case is dealing with the transient mixing of a clear plug in the vessel when one primary pump starts-up. We compare the numerical and experiment results giving the mean boron concentration at the core inlet for several clear water plugs. The results show again a good agreement. (authors). 12 refs., 10 figs., 1 tab

  18. Effect of Ni and Cr on IGSCC growth rate of Ni-Cr-Fe alloys in PWR primary water

    International Nuclear Information System (INIS)

    Arioka, K.; Yamada, T.; Aoki, M.; Miyamoto, T.

    2015-01-01

    The purpose of this research is to examine the dependence of SCC (Stress Corrosion Crack) growth on nickel and chromium in PWR primary water; the objective is to obtain the basic knowledge to understand SCC behavior of steam generator tubing materials. The second objective is to understand whether accelerated testing at higher temperatures is appropriate for predicting SCC initiation and growth at lower temperatures. For these objectives, SCC growth was measured in PWR primary water at 290, 320, 330, 340, and 360 C. degrees under static load conditions. Tests were performed using 0.5 T compact tension type specimen using 20%CW X%Ni-16%Cr-Fe alloys in the range of nickel concentration between 16 to 60% and laboratory melted nuclear grade 20% cold worked Alloy 800 (USN N08800, CW800NG). Four important patterns were observed. First, significant effect of nickel on IGSCC resistance was observed at 340 and 360 C. degrees. The rate of IGSCC growth decreases with increasing nickel concentration in the range of nickel concentration between 10% to 25% nickel; and then, the rate of IGSCC increases with increasing nickel concentration in the range of Ni content between 50% and 76%. This trend is quite similar to the results reported by Coriou and Staehle tested in deaerated pure water at 350 C. degrees. However, no significant dependence of Ni content on IGSCC in PWR water at 320 and 290 C. degrees was observed. The change in SCC growth dependence on nickel concentration suggested that the main rate limiting processes on IGSCC growth seems to change between 320 and 340 C. degrees. Secondly, significant beneficial effects of chromium in alloys were observed at 320 C. degrees. However, no beneficial effect of chromium addition in alloys was observed at 360 C. degrees. Thirdly, peak temperatures in growth rate of IGSCC were observed in almost all test materials except for 20%CW Alloy 600. Finally, intergranular attack was observed in some alloys at lower temperature, and the

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

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Junjie; Xiao, Qian [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); State Key Laboratory of Advanced Special Steels, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Lu, Zhanpeng, E-mail: zplu@t.shu.edu.cn [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); State Key Laboratory of Advanced Special Steels, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Shanghai Key Laboratory of Advanced Ferrometallurgy, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Ru, Xiangkun; Peng, Hao; Xiong, Qi; Li, Hongjuan [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China)

    2017-06-15

    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. - Highlights: •Long-term EIS measurements of 316L SS in simulated PWR primary water. •Highest charge-transfer resistance and oxide film resistance in oxygenated water. •Highest electric double-layer capacitance and oxide film CPE in hydrogenated water. •Similar compositions, different shapes of oxides in deaerated/hydrogenated water. •Inner layer Cr-rich in hydrogenated/deaerated water, Ni-rich in oxygenated water.

  20. Oligo-cyclic damage and behaviour of a 304 L austenitic stainless steel according to environment (vacuum, air, PWR primary water) at 300 C

    International Nuclear Information System (INIS)

    De Baglion, L.

    2011-01-01

    Nowadays, for nuclear power plants licensing or operating life extensions, various safety authorities require the consideration of the primary water environment effect on the fatigue life of Pressurized Water Reactor (PWR) components. Thus, this work focused on the study of low cycle fatigue damage kinetics and mechanisms, of a type 304L austenitic stainless steel. Several parameters effects such as temperature, strain rate or strain amplitude were investigated in air as in PWR water. Thanks to targeted in-vacuum tests, the intrinsic influence of these parameters and environments on the fatigue behaviour of the material was studied. It appears that compared with vacuum, air is already an active environment which is responsible for a strong decrease in fatigue lifetime of this steel, especially at 300 C and low strain amplitude. The PWR water coolant environment is more active than air and leads to increased damage kinetics, without any modifications of the initiation sites or propagation modes. Moreover, the decreased fatigue life in PWR water is essentially attributed to an enhancement of both initiation and micropropagation of 'short cracks'. Finally, the deleterious influence of low strain rates on the 304L austenitic stainless steel fatigue lifetime was observed in PWR water environment, in air and also in vacuum without any environmental effects. This intrinsic strain rate effect is attributed to the occurrence of the Dynamic Strain Aging phenomenon which is responsible for a change in deformation modes and for an enhancement of cracks initiation. (author)

  1. Quantitative measurement of trace amounts of dissolved oxygen in the primary and secondary systems of PWR nuclear power plants

    International Nuclear Information System (INIS)

    Castaneda, H.B.; Neale, T.A.

    1989-01-01

    Establishing and maintaining the correct water chemistry conditions in the primary and secondary systems of pressurized water reactor (PWR) nuclear power plants is essential in order to maximize the operating life and guarantee the uninterrupted availability of the major components of each PWR unit. The exact specifications for maintaining the correct water chemistry are well established. One of the most important parameters that must be closely monitored in a modern power generation plant is the level of dissolved oxygen (DO) present in the system. Because of the high temperatures and pressures involved, even minute traces of DO---on the order of a few parts per billion (ppb)---can be detrimental to the heat transfer surfaces in steam generators, heaters, etc. The authors argue that the method of determining trace levels of DO presented here is a modification of the original method that has greatly increased the detection level obtainable with Rhodazine-D. Measurements down to less than 1 ppb (μg/Liter), with a resolution of 0.5 ppb (μ/Liter), are now easily obtainable. No calibration procedures are required and no maintenance of critical components is needed. This quantitative method is based on the instantaneous stoichiometric reaction of Rhodazine-D with oxygen. After less than one minute the oxidation reaction is complete and the fully developed color is compared with a set of stable liquid color standards. The color standards are formulated using the oxidized form of Rhodazine-D, thus providing an exact color match for the reacted sample-reagent. Supporting data are presented that confirm the relative accuracy and sensitivity of the new method, as well as results of a comparative evaluation of the method versus in-line dissolved oxygen analyzers

  2. Measurement and Elemental Analysis of Muria Sea Water for Primary Water of PWR

    International Nuclear Information System (INIS)

    Dwi Biyantoro; Kris Tri Basuki

    2007-01-01

    Treatment process of water of free mineral and study area of processing of sea water to meet the need of water cooling of PWR. Desalination is a separation process used to reduce the dissolved salt content of saline water. All desalination processes involve three liquid streams: the saline feedwater (brackish water or seawater), low salinity product water, and very saline concentrate. Seawater Muria Jepara is feed type 1 with content salt is low and relative suspension and weight metals is low. There are two types of membrane process used for desalination: reverse osmosis (RO) and electrodialysis (ED). The starting of process is difficult and expensive. However, starting since 1950, desalination process appear to be economically practical for ordinary use. There are 3 key elements significantly affect to the technical performance and long term characteristic type, i.e. 1) energy, 2) corrosivity of seawater, and 3) desalination process technology. These elements closely inter related and important for design improvements and technical performance optimization. From the analysis of sea water of gulf Muria Jepara was found as follows : Fe = 0.176 ± 0.012 ppm, Pb = 0.612 ± 0.017 ppm, Cd = 52.567 ± 0.750 ppm, Cu = 0.044 ± 0.005 ppm, Zn = 0.061 ± 0.003 ppm, Mn = 0.057 ± 0.003 ppm, Ca between = 365.256 - 368.654 ppm, Na between = 9572.000 - 9775.000 ppm, Mg between 759.000 - 779.00 ppm and Ni = 0.524 ± 0.005 ppm. Cost production of process using reverse osmosis as around 0.9 - 1 US$/m3, while using electrodialysis is around 1.2 US$/m3, and by using evaporation process or distillation process is around 1.4 - 1.6 US$/m3. (author)

  3. Advancing Continence in Typically Developing Children: Adapting the Procedures of Foxx and Azrin for Primary Care.

    Science.gov (United States)

    Warzak, William J; Forcino, Stacy S; Sanberg, Sela Ann; Gross, Amy C

    2016-01-01

    To (1) identify and summarize procedures of Foxx and Azrin's classic toilet training protocol that continue to be used in training typically developing children and (2) adapt recent findings with the original Foxx and Azrin procedures to inform practical suggestions for the rapid toilet training of typically developing children in the primary care setting. Literature searches of PsychINFO and MEDLINE databases used the search terms "(toilet* OR potty* AND train*)." Selection criteria were only peer-reviewed experimental articles that evaluated intensive toilet training with typically developing children. Exclusion criteria were (1) nonpeer reviewed research, (2) studies addressing encopresis and/or enuresis, (3) studies excluding typically developing children, and (4) studies evaluating toilet training during infancy. In addition to the study of Foxx and Azrin, only 4 publications met the above criteria. Toilet training procedures from each article were reviewed to determine which toilet training methods were similar to components described by Foxx and Azrin. Common training elements include increasing the frequency of learning opportunities through fluid loading and having differential consequences for being dry versus being wet and for voiding in the toilet versus elsewhere. There is little research on intensive toilet training of typically developing children. Practice sits and positive reinforcement for voids in the toilet are commonplace, consistent with the Foxx and Azrin protocol, whereas positive practice as a corrective procedure for wetting accidents often is omitted. Fluid loading and differential consequences for being dry versus being wet and for voiding in the toilet also are suggested procedures, consistent with the Foxx and Azrin protocol.

  4. DRUFAN-01/MOD2, Transient Thermohydraulics of PWR Primary System LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Burwell, M J; Lerchl, G; Steinhoff, F; Wolfert, K [Gesellschaft fuer Reaktorsicherheit (GRS) mbH, Forschungsgelaende, 8046 Garching (Germany)

    1982-12-13

    1 - Description of problem or function: DRUFAN is an advanced best estimate code for simulation of the transient thermal hydraulic behaviour during PWR-blowdown with large break size. 2 - Method of solution: The code is based on the lumped parameter approach and allows flexible control volume configurations. The physical model takes into account thermodynamic nonequilibrium. Using finite difference techniques a 1-dimensional representation of the discharge flow path including geometrical influences is possible. The physical model is based on separated field equations for liquid and vapour mass and overall field equations for energy and momentum. The mass transfer rates between phases during evaporation and condensation are based on correlations for the controlled growth and shrinkage of vapour bubbles or liquid droplets, respectively. A heat conductor model based on the energy transport equation is available for simulation of structures, electrical heater rods and fuel rods. For the heat transfer between solid structures and the fluid a comprehensive package of flow regime dependent heat transfer and critical heat flux correlations can be used. Simulation of components (valve, pressurizer, accumulator, pump, steam generator) is possible with functions or models. Power generation in solid structures may be simulated by an input time function, an electrical heater model or a neutron kinetics models. As a result of the lumped parameter approach a set of ordinary differential equations is obtained from the field equations. These equations, together with those resulting from the simulation of critical discharge flow near the outlet by a finite difference method, are solved by an explicit/implicit integration method with automatic time step, order and error control. The ordinary differential equations representing heat conductors are solved by an essentially implicit integration method. 3 - Restrictions on the complexity of the problem: - Vapour or liquid phase are

  5. Volumetric and chemical control auxiliary circuit for a PWR primary circuit

    International Nuclear Information System (INIS)

    Costes, D.

    1990-01-01

    The volumetric and chemical control circuit has an expansion tank with at least one water-steam chamber connected to the primary circuit by a sampling pipe and a reinjection pipe. The sampling pipe feeds jet pumps controlled by valves. An action on these valves and pumps regulates the volume of the water in the primary circuit. A safety pipe controlled by a flap automatically injects water from the chamber into the primary circuit in case of ruptures. The auxiliary circuit has also systems for purifying the water and controlling the boric acid and hydrogen content [fr

  6. Prevention against fragile fracture in PWR pressure vessel in the presence of pressurized thermal shock

    International Nuclear Information System (INIS)

    Carmo, E.G.D. do; Oliveira, L.F.S. de; Roberty, N.C.

    1984-01-01

    A method for the determination of operational limit curves (primary pressure versus temperature) for PWR is presented. Such curves give the operators indications related to the safety status of the plant concerning the possibility of a pressurized thermal shock. The method begins by a thermal analysis for several postulated transients, followed by the determination of the thermomechanical stresses in the vessel and finally it makes use of the linear elasticity fracture mechanics. Curves are shown for a typical PWR. (Author) [pt

  7. Evaluation of the fuel rod integrity in PWR reactors from the spectrometric analysis of the primary coolant

    International Nuclear Information System (INIS)

    Monteiro, Iara Arraes

    1999-02-01

    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)

  8. Reliability of an HTR-module primary circuit pressure boundary. Influences, sensitivity, and comparison with a PWR

    International Nuclear Information System (INIS)

    Staat, M.

    1995-01-01

    The reliability of the HTR-module for electricity and steam generation was analysed for normal operation, as well as accident conditions. The probabilistic fracture mechanics assessment was performed with a modification of the ZERBERUS code on the basis of widely used data. The calculated failure probabilities may thus be compared with similar investigations. The HTR-module primary circuit pressure boundary as a unit showed leak-before-break behaviour in a probabilistic sense, although a break was more probable than a leak for some of its parts.However, the findings may depend greatly on the stochastic data. Therefore a stochastic reference problem is defined and the results are compared with the Japanese round robin on a PWR section. Possible changes of failure probabilities and of the leak-before-break behaviour are discussed for different criteria for the events leading to a leak, and for modifications of the stochastic reference problem such as the inclusion of NDE. The results may be used to identify those stochastic variables which have the greatest influence on the computed failure probabilities, and to perhaps justify further work which would provide more detailed information on these probabilities. Furthermore, there is an obvious need for reduction of the non-statistical reasons for great variations of failure probabilities. (orig.)

  9. Quantitative assessment of intergranular damage due to PWR primary water exposure in structural Ni-based alloys

    International Nuclear Information System (INIS)

    Ter-Ovanessian, Benoît; Deleume, Julien; Cloué, Jean-Marc; Andrieu, Eric

    2013-01-01

    Highlights: ► IG damage occurred on Ni-base alloys during exposure at high temperature water. ► Two characterization methods yield a tomographic analysis of this IG damage. ► Connected or isolated intergranular oxygen/oxide penetrations are quantified. ► Such quantitative description provides information on IGSCC susceptibility. - Abstract: Two nickel-based alloys, alloy 718 and alloy 600, known to have different resistances to IGSCC, were exposed to a simulated PWR primary water environment at 360 °C for 1000 h. The intergranular oxidation damage was analyzed in detail using an original approach involving two characterization methods (Incremental Mechanical Polishing/Microcopy procedure and SIMS imaging) which yielded a tomographic analysis of the damage. Intergranular oxygen/oxide penetrations occurred either as connected or isolated penetrations deep under the external oxide/substrate interface as far as 10 μm for alloy 600 and only 4 μm for alloy 718. Therefore, assessing this damage precisely is essential to interpret IGSCC susceptibility.

  10. Effects of PbO on the oxide films of incoloy 800HT in simulated primary circuit of PWR

    International Nuclear Information System (INIS)

    Tan, Yu; Yang, Junhan; Wang, Wanwan; Shi, Rongxue; Liang, Kexin; Zhang, Shenghan

    2016-01-01

    Effects of trace PbO on oxide films of Incoloy 800HT were investigated in simulated primary circuit water chemistry of PWR, also with proper Co addition. The trace PbO addition in high temperature water blocked the protective spinel oxides formation of the oxide films of Incoloy 800HT. XPS results indicated that the lead, added as PbO into the high temperature water, shows not only +2 valance but also +4 and 0 valances in the oxide film of 800HT co-operated with Fe, Cr and Ni to form oxides films. Potentiodynamic polarization results indicated that as PbO concentration increased, the current densities of the less protective oxide films of Incoloy 800HT decreased in a buffer solution tested at room temperature. The capacitance results indicated that the donor densities of oxidation film of Incoloy 800HT decreased as trace PbO addition into the high temperature water. - Highlights: • Trace PbO addition into the high temperature water block the formation of spinel oxides on Incoloy 800HT. • The donor density of oxide film decreases with trace PbO addition. • The current density of potentiodynamic polarization decreases of oxide film with trace PbO addition.

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

    Science.gov (United States)

    Ming, Hongliang; Zhang, Zhiming; Wang, Jiazhen; Zhu, Ruolin; Ding, Jie; Wang, Jianqiu; Han, En-Hou; Ke, Wei

    2015-05-01

    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 (Fe3O4, Fe2O3 and a small amount of NiFe2O4, Ni(OH)2, Cr(OH)3 and (Ni, Fe)Cr2O4) and a Cr-rich inner layer (FeCr2O4 and NiCr2O4) 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.

  12. PWR core design calculations

    International Nuclear Information System (INIS)

    Trkov, A.; Ravnik, M.; Zeleznik, N.

    1992-01-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [sl

  13. T/sub hot/ reduction: a program for lowering primary temperatures on a PWR

    International Nuclear Information System (INIS)

    Augustine, D.B.; DiTommaso, S.M.; Manz, E.M.; Reister, P.

    1987-01-01

    This paper focuses on the key technical issues addressed in a program to support operation of the Byron Unit 1 pressurized water reactor at primary side temperatures significantly lowered with respect at primary side temperatures significantly lowered with respect to the original design temperatures. These operating temperatures were lowered in order to reduce the potential for initiation of primary water stress corrosion cracking in the steam generator tubing. The efforts of this program were aimed at maintaining operation of the unit at the maximum possible power level at the reduced temperatures. In addition, the program is designed to allow for cycle-to-cycle flexibility within a range of operating temperatures from the original design temperatures to temperatures lowered by ∼ 11 0 C (20 0 F)

  14. Study of colloidal particles behaviour in the PWR primary circuit conditions

    International Nuclear Information System (INIS)

    Barale, M.

    2006-12-01

    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)

  15. Integrated equipment for increasing and maintaining coolant pressure in primary circuit of PWR nuclear power plant

    International Nuclear Information System (INIS)

    Sykora, D.

    1986-01-01

    An open heat pump circuit is claimed connected to the primary circuit. The pump circuit consists of a steam pressurizer with a built-in steam distributor, a compressor, an expander, a reducing valve, an auxiliary pump, and of water and steam pipes. The operation is described and a block diagram is shown of integrated equipment for increasing and maintaining pressure in the nuclear power plant primary circuit. The appropriate entropy diagram is also shown. The advantage of the open pump circuit consists in reducing the electric power input and electric power consumption for the steam pressurizers, removing entropy loss in heat transfer with high temperature gradient, in the possibility of inserting, between the expander and the auxiliary pump, a primary circuit coolant treatment station, in simplified design and manufacture of the high-pressure steam pressurizer vessel, reducing the weight of the steam pressurizer by changing its shape from cylindrical to spherical, increasing the rate of pressure growth in the primary circuit. (E.S.)

  16. MELCOR 1.8.3 application to NUPEC M-7-1 test (ISP-35) and two hydrogen severe accident scenarios in a typical PWR plant

    International Nuclear Information System (INIS)

    Jimenez Garcia, M.A.; Martin-Fuertes, F.; Martin-Valdepenas, J.M.

    1997-01-01

    Combustion of the hydrogen released to the containment during a severe accident is one of the issues to establish the real threats to the third barrier integrity in nuclear power facilities. Computational efforts on management procedures, such as the containment spray operation, are being addressed at the CTN-UPM to cope with the problem. On top of this, studies about in-containment hydrogen distribution and combustion are currently carried out with the codes MELCOR 1.8.3 and ESTER 1.0-RALOC 2.2. In this study, MELCOR 1.8.3 has been validated against the NUPEC M-7-1 Test, which already showed in 1993 that a good agreement was reached out when the previous MELCOR 1.8.2 calculations were performed regarding to the helium distribution throughout the facility. Nevertheless, some discrepancies were detected when analysing wall and atmosphere temperatures. Generally, well-mixed atmosphere scenarios, in which the role played by the containment water spraying is of the major importance, appear when such a mechanism promotes the onset of convection driven flow patterns that rapidly homogenize the gas properties. The purpose of the new MELCOR 1.8.3 assessment is to take advantage of the newest implemented models to obtain a more realistic thermalhydraulics simulation. A variation case was also performed to highlight the influence of water spray operation. In a second part of the study, insights coming from the previous work were used to apply MELCOR 1.8.3 models to a SBO severe accident scenario management in a commercial 2700 MWt 3-loop W PWR containment

  17. Modeling PWR systems for monitoring primary-to-secondary leakage using tritium tracer

    International Nuclear Information System (INIS)

    Peiffer, D.G.

    1992-01-01

    This paper discusses several techniques available for monitoring primary to secondary leakage, focusing on the advantages and disadvantages of each. A mathematical model of Millstone 2 describes the behavior of tritium activity in the secondary plant water when a leak exists. Real data from Millstone 2 illustrate the accuracy and reliability of the model and use of the model to measure the mass of water in the secondary system

  18. Recent bibliography on analytical and sampling problems of a PWR primary coolant

    International Nuclear Information System (INIS)

    Illy, H.

    1980-07-01

    An extensive bibliography on the problems of analysis and sampling of the primary cooling water of PWRs is presented. The aim was to collect the analytical methods for dissolved gases. The sampling and preparation are also taken into account. last 8-10 years is included. The bibliography is arranged into alphabetical order by topics. The most important topics are as follows: boric acid, gas analysis, hydrogen isotopes, iodine, noble gases, radiation monitoring, sampling and preparation, water chemistry. (R.J.)

  19. PWR type reactor equipped with a primary circuit loop water level gauge

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro.

    1990-01-01

    The time of lowering a water level to less than the position of high temperature side pipeway nozzle has been rather delayed because of the swelling of mixed water level due to heat generation of the reactor core. Further, there has been a certain restriction for the installation, maintenance and adjustment of a water level gauge since it is at a position under high radiation exposure. Then, a differential pressure type water level gauge with temperature compensation is disposed at a portion below a water level gauge of a pressurizer and between the steam generator exit plenum and the lower end of the loop seal. Further, a similar water level system is disposed to all of the loops of the primary circulation circuits. In a case that the amount of water contained in a reactor container should decreased upon occurrence of loss of coolant accidents caused by small rupture and stoppage of primary circuit pumps, lowering of the water level preceding to the lowering of the water level in the reactor core is detected to ensure the amount of water. Since they are disposed to all of the loops and ensure the excess margin, reliability for the detection of the amount of contained water can be improved by averaging time for the data of the water level and averaging the entire systems, even when there are vibrations in the fluid or pressure in the primary circuit. (N.H.)

  20. Contamination of a PWR primary circuit by fuel pins with failed cladding

    International Nuclear Information System (INIS)

    Janvier, J.C.; Chagrot, M.

    1979-01-01

    The safety authorities in the principal nuclear countries appear to be attaching increasing importance to keeping reactor primary circuits as contamination-free as possible. Therefore, the consequences of cladding failures and especially of those resulting from fabrication defects have to be evaluated, for when these failures become systematic in nature they constitute an important source of contamination in pressurized-water reactors. The Grenoble Nuclear Research Centre is implementing a programme on the study of such failures with a view to analysing the behaviour of failed fuel elements. A distinction is made between two types of cladding failure, depending on whether the primary water enters the fuel pin as soon as the circuits are pressurized (fabrication defect) or whether the failure is caused during operation. The emission of gaseous fission products and halogens has been analysed in different operating modes (steady-state or transient), and in spite of the complexity of the phenomena involved, some results have been obtained which already enable one to evaluate fission product contamination of the primary circuit. (author)

  1. Effect of strain-path on stress corrosion cracking of AISI 304L stainless steel in PWR primary environment at 360 deg. C

    International Nuclear Information System (INIS)

    Couvant, T.; Vaillant, F.; Boursier, JM.; Delafosse, D.

    2004-01-01

    Austenitic stainless steels (ASS) are widespread in primary and auxiliary circuits of PWR. Moreover, some components suffer stress corrosion cracking (SCC) under neutron irradiation. This degradation could be the result of the increase of hardness or the modification of chemical composition at the grain boundary by irradiation. In order to avoid complex and costly corrosion facilities, the effects of irradiation on the material are commonly simulated by applying a cold work on non-irradiated material prior to stress corrosion cracking tests. Slow strain rate tests were conducted on an austenitic stainless steel (SS) AISI 304L in PWR environment (360 deg. C). Particular attention was directed towards pre-straining effects on crack growth rate (CGR) and crack growth path (CGP). Results have demonstrated that the susceptibility of 304L to SCC in high-temperature hydrogenated water was enhanced by pre-straining. It seemed that IGSCC was enhanced by complex strain paths. (authors)

  2. Implementation of the structural integrity analysis for PWR primary components and piping

    International Nuclear Information System (INIS)

    Pellissier-Tanon, A.

    1982-01-01

    The trends on the definition, the assessment and the application of fracture strength evaluation methodology, which have arisen through experience in the design, construction and operation of French 900-MW plants are reviewed. The main features of the methodology proposed in a draft of Appendix ZG of the RCC-M code of practice for the design verification of fracture strength of primary components are presented. The research programs are surveyed and discussed from four viewpoints, first implementation of the LEFM analysis, secondly implementation of the fatigue crack propagation analysis, thirdly analysis of vessel integrity during emergency core cooling, and fourthly methodology for tear fracture analysis. (author)

  3. Identification of Standing Pressure Waves Sources in Primary Loops of NPP with WWER and PWR

    Directory of Open Access Journals (Sweden)

    K.N. Proskuriakov

    2016-05-01

    Full Text Available Results of measurement and calculation of Eigen frequencies of coolant pressure oscillations in primary loops of NPP are presented. The simple calculation model based on equivalence of electric circuit with elastic wave propagation in liquids and gases, which gives a sensible interpretation of standing pressure waves sources is developed. It is shown, that pressurizer manifest itself as managed Helmholtz resonator generating a number of SPW (with Eigen frequencies of steam volume, water volume and their combination with coolant volume of respiratory line.

  4. High temperature filtration of radioactivable corrosion products in the primary circuit of PWR type reactors

    International Nuclear Information System (INIS)

    Dolle, L.

    1976-01-01

    A effective limitation to the deposition of radioactive corrosion products in the core of a reactor at power operation, is to be obtained by filtering the water of the primary circuit at a flow rate upper than 1% of the coolant flow rate. However, in view of accounting for more important release of corrosion products during the reactor start-up and also for some possible variations in the efficiency of the system, it is better that the flow rate to be treated by the cleaning circuit is stated at 5%. Filtration must be effected at the temperature of the primary circuit and preferably on each loop. To this end, the feasibility of electromagnetic filtration or filtration through a deep bed of granulated graphite has been studied. The on-loop tests effected on each filter gave efficiencies and yields respectively upper than 90% and 99% for magnetite and ferrite particles in suspension in water at 250 deg C. Such results confirm the interest lying in high temperature filtration and lead to envisage its application to reactors [fr

  5. Study of the corrosion products in the primary system of PWR plants as the source of radiation fields build-up

    International Nuclear Information System (INIS)

    Brabant, R. van; Regge, P. de.

    1982-01-01

    In the first part the behaviour of the corrosion products in the primary system of PWR plants is depicted on the basis of a literature review of the field. Water chemistry, corrosion processes and activation of corrosion products are the main topics. In the second part the results of the characterization of corrosion particles in the primary coolant circuit of the Doel 1 and 2 reactors are described, during steady state operation and transient phases. In the third part the possibilities for radiation control at nuclear power plants are outlined. The filtration possibilities for the reactor coolant are explored in detail. (author)

  6. Control of the flanges of the thermal barriers fitting the 900 MWe PWR primary pumps

    International Nuclear Information System (INIS)

    Cleurennec, M.; Thebault, Y.; Abittan, E.; Pages, C.; Lhote, P.A.; Randrianarivo, L.

    1998-01-01

    During maintenance visit on 93 D type primary pumps of French 900 MWe nuclear units, cracking has been evidenced on the thermal barrier, first on the flange, on the face of connection of the cooling, water coils, and then on the weld between the housing and the flange. Laboratory examinations have exhibited that this cracking is due to a fatigue phenomenon which is initiated on locations where high residual stresses are present. One pump, in service in a plant, has received an instrumentation in order to determine stress cycling. Measurements of temperature on the surface of the metal have shown the presence of thermal cycling due to the thermohydraulic conditions inside the thermal barrier. A non destructive testing method using ultrasounds has been developed in order to asses the magnitude cracking. Corrective and preventive actions have been implemented for repairing and improving thermal barrier when cracking is detected. (authors)

  7. Effect of Residual Strain on PWSCC Initiation of Alloy 690 in Primary Water of PWR

    International Nuclear Information System (INIS)

    Lee, Eun-Hee; Kim, Sung-Woo; Eom, Ki-Hyeon; Hwang, Seong-Sik

    2016-01-01

    The object of this study is to attempt to correlate the susceptibility to crack initiation of Alloy 690 with the levels of cold work in simulated primary water. Experiments were conducted in high-temperature water to accelerate the cracking behavior and to determine the severity of crack initiation as a function of level of cold work for Alloy 690. Cracking susceptibility was determined by measuring the crack length per unit area and crack density. SCC initiation susceptibility of Alloy 690 increased with increase of the residual strain induced by the prior cold-rolling process. The 20% cold-rolled specimen exhibited mostly IG cracks, while the 30% and 40% cold-rolled specimens revealed significant amounts of IG and TG cracks. Most cracks were observed near the smallest cross-section area of the tapered specimen, where the maximum stress was applied, indicating that the SCC initiation susceptibility was strongly dependent on the applied stress. In order to find the correlation between SCC initiation susceptibility and residual strain induced by cold work, more analyses need to be performed in terms of the crack length per unit area and the crack density, as a strong indicative for SCC initiation susceptibility

  8. Importance of ECP in the prediction of radiation fields in PWR and VVER primary circuits

    International Nuclear Information System (INIS)

    Urquidi-Macdonald, M.; Jacesko, S.L.; Macdonald, Digby D.; Salter-Williams, M.

    2002-01-01

    A model has been developed for predicting mass and activity transport in the primary coolant circuits of PWRs and VVERs with the objective of demonstrating and quantifying the importance of the electrochemical corrosion potential (ECP) in determining the impact of both processes on reactor operation. The model initially employs a radiolysis/mixed potential code to calculate the ECP at four locations (core, hot leg, steam generator, cold leg) and the ECP is then used to estimate the local magnetite solubility. The solubility is then averaged around the loop to yield the ''background'' solubility. Comparison of the background solubility with the local solubility determines whether precipitation or dissolution will occur at any given point in the circuit under any given set of conditions. It is further assumed that the concentration of 59 Co in the coolant is given by the isotopic fraction of this species compared with iron averaged over all materials and weighted by the respective wetted areas. Activation of 59 Co to 60 Co is assumed to occur in the coolant phase by fast, epithermal, and thermal neutron capture. The calculated activity is then used to train an artificial neural network (ANN) to establish relationships between activity at any given location and the operating properties of the reactor, including coolant pH, ECP, temperature, power level, etc. The model predicts that during shut down, magnetite (and hence 59 Co) migrates to the core, where it is irradiated and activated, particularly during subsequent start-up. During start-up, the magnetite (and hence 60 Co) migrates from the core to out-of-core surfaces where it establishes the radiation fields. (authors)

  9. PWR primary system chemistry: Experience with elevated pH at Millstone Point Unit 3

    International Nuclear Information System (INIS)

    Bergmann, C.A.; McAtee, K.R.; Roesmer, J.

    1990-08-01

    The maintenance of an elevated pH of the hot reactor coolant in several plants for a fraction of their respective fuel cycles confirmed earlier laboratory findings that an elevated pH (>6.9 at temperature) resulted in reduced component radiation fields. In view of the potential radiological benefits of elevated coolant pH operation, Northeast Utilities (NU) in support of an EPRI- Westinghouse program, agreed to operate the Millstone 3 plants as a demonstration of elevated coolant pH during operation with extended fuel cycles. This report represents the data and describes the effects of operation of the Millstone 3 cycle 2 core at an elevated hot coolant pH of 7.0--7.5 on Zircaloy cladding corrosion, and on the buildup of core crud and excore radiation fields. The following preliminary conclusion were made based on an evaluation of the data obtained after one cycle of Millstone 3 elevated coolant pH operation. Although some of the measured zirconium dioxide thicknesses are greater than expected, the data does not indicate a significant lithium effect on the corrosion rate of the fuel cladding because of the relatively high variability of thickness measurements on rods of similar duty. There was no observable effect of the higher pH operation on any of the other fuel assembly components. Changes in core crude deposit thickness, chemical and radiochemical composition and apparent residence time were noted. These changes are consistent with the expected effects of a more dissolving coolant on activity and crud transport in a primary circuit. The overall radiation exposure rate trends from cycle 1 to cycle 2 indicated a favorable effect of the higher pH operation. 17 figs., 11 tabs

  10. Oxidation of Alloy 82 in nominal PWR primary water at 340 deg. C and in hydrogenated steam at 400 deg. C

    International Nuclear Information System (INIS)

    Chaumun, Elizabeth; Guerre Catherine; Duhamel, Cecilie; Sennour, Mohamed; Curieres, Ian-de

    2012-09-01

    Nickel-base weld metals are susceptible to stress corrosion cracking (SCC) in Pressurized Water Reactor (PWR) primary water. As tests in laboratory need to last, in some cases, at least several thousand hours to get stress corrosion crack initiation or propagation in simulated primary water, pure hydrogenated steam at 400 deg. C was used to perform accelerated tests. To confirm that these conditions are still representative of primary water conditions, results of oxidation tests of coupons in hydrogenated steam at 400 deg. C and in primary water at 340 deg. C have been compared. Surface oxide layers have been characterized in order to discuss the influence of the temperature and of the media (water or steam). (authors)

  11. The hold-time effects on the low cycle fatigue behaviors of 316 SS in PWR primary environment

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Junho; Hong, Jong-Dae; Seo, Myung-Gyu; Jang, Changheui [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    The effects of the environments on fatigue life of the structural materials used in nuclear power plants (NPPs) were known to be significant according to the extensive test results. Accordingly, the fatigue analysis procedures and the design fatigue curves were proposed in the ASME Code. However, the implication that the existing ASME design fatigue curves did not sufficiently reflect the effect of the operation conditions of nuclear power plants emerged as an issue to be resolved. One of possible reasons to explain the discrepancy is that the laboratory test conditions do not represent the actual plant transients. Therefore, it is necessary to clarify the effects of light water environments on fatigue life while considering more plant-relevant transient conditions such as hold-time. For this reason, this study will focus on the fatigue life of type 316 stainless steel (SS) in the pressurized water reactor (PWR) environments while incorporating the hold-time during the low cycle fatigue (LCF) test in simulated PWR environments. The objective of this study is to characterize the effects of hold-time 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 SS in 310 .deg. C air and simulated PWR environments. To simulate the heat-up and cool-down transient, sub-peak strain holding during the down-hill of strain amplitude was chosen. Currently, LCF tests with 60 seconds holding are in progress. The 0.4, 0.04%/s strain rate condition test results are presented in this study, which shows somewhat longer fatigue life.

  12. The hold-time effects on the low cycle fatigue behaviors of 316 SS in PWR primary environment

    International Nuclear Information System (INIS)

    Lee, Junho; Hong, Jong-Dae; Seo, Myung-Gyu; Jang, Changheui

    2015-01-01

    The effects of the environments on fatigue life of the structural materials used in nuclear power plants (NPPs) were known to be significant according to the extensive test results. Accordingly, the fatigue analysis procedures and the design fatigue curves were proposed in the ASME Code. However, the implication that the existing ASME design fatigue curves did not sufficiently reflect the effect of the operation conditions of nuclear power plants emerged as an issue to be resolved. One of possible reasons to explain the discrepancy is that the laboratory test conditions do not represent the actual plant transients. Therefore, it is necessary to clarify the effects of light water environments on fatigue life while considering more plant-relevant transient conditions such as hold-time. For this reason, this study will focus on the fatigue life of type 316 stainless steel (SS) in the pressurized water reactor (PWR) environments while incorporating the hold-time during the low cycle fatigue (LCF) test in simulated PWR environments. The objective of this study is to characterize the effects of hold-time 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 SS in 310 .deg. C air and simulated PWR environments. To simulate the heat-up and cool-down transient, sub-peak strain holding during the down-hill of strain amplitude was chosen. Currently, LCF tests with 60 seconds holding are in progress. The 0.4, 0.04%/s strain rate condition test results are presented in this study, which shows somewhat longer fatigue life

  13. The effect of zinc injection into PWR primary coolant on the reduction of radiation buildup and corrosion control. The solubilities of zinc, nickel and cobalt spinel oxides

    International Nuclear Information System (INIS)

    Miyajima, Kaori; Hirano, Hideo

    1999-01-01

    The use of zinc injection into PWR primary coolant to reduce radiation buildup has been widely studied, and te reduction effect has been experimentally confirmed. However, some items, such as the optimal concentration of zinc required to reduce radiation buildup, the corrosion control effect of zinc injection, and the influence of zinc injection on the integrity of fuel cladding, have not been clarified yet. In particular, the corrosion suppression effect of zinc remains unconfirmed. Therefore, it is necessary to measure and calculate the solubilities of zinc and nickel spinel oxides, which are formed on the surface of Ni-based alloys in PWR primary systems. In this study, in order to assess the effectiveness of zinc injection in the reduction of radiation buildup and the corrosion control of Ni-based alloy, the potential-pH diagrams for Zn-Cr-H 2 O, Ni-Cr-H 2 O, and Co-Cr-H 2 O systems at 300degC were constructed and the solubilities of Zn-Cr, Ni-Cr, and Co-Cr spinel oxides were calculated. It is concluded that under pH conditions for which NiCr 2 O 4 is stable, zinc injection is effective in corrosion control as well as in reducing radiation buildup. (author)

  14. 3. Workshop for IAEA ICSP on Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Containment and Primary System during Accidents. Presentations

    International Nuclear Information System (INIS)

    2012-04-01

    Most advanced nuclear power plant designs adopted several kinds of passive systems. Natural circulation is used as a key driving force for many passive systems and even for core heat removal during normal operation such as NuScale, CAREM, ESBWR and Indian AHWR designs. Simulation of natural circulation phenomena is very challenging since the driving force of it is weak compared to forced circulation and involves a coupling between primary system and containment for integral type reactor. The IAEA ICSP (International Collaborative Standard Problem) on 'Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Containment and Primary System during Accidents' was proposed within the CRP on 'Natural Circulation Phenomena, Modelling, and Reliability of Passive Systems that utilize Natural Circulation'. Oregon State University (OSU) of USA offered to host this ICSP. This ICSP plans to conduct the following experiments and blind/open simulations with system codes: 1. Quasi-steady state operation with different core power levels: Conduct quasi-steady state operation with step-wise increase of core power level in order to observe single phase natural circulation flow according to power level. The experimental facility and operating conditions for an integral PWR will be used. 2. Thermo-hydraulic Coupling between Primary system and Containment: Conduct a loss of feedwater transient with subsequent ADS blowdown and long term cooling to determine the progression of a loss of feedwater transient by natural circulation through primary and containment systems. These tests would examine the blowdown phase as well as the long term cooling using sump natural circulation by coupling the primary to containment systems. This data could be used for the evaluation of system codes to determine if they model specific phenomena in an accurate manner. OSU completed planned two ICSP tests in July 2011 and real initial and boundary conditions measured from the

  15. Patients with primary insomnia in the sleep laboratory: do they present with typical nights of sleep?

    Science.gov (United States)

    Hirscher, Verena; Unbehaun, Thomas; Feige, Bernd; Nissen, Christoph; Riemann, Dieter; Spiegelhalder, Kai

    2015-08-01

    The validity of sleep laboratory investigations in patients with insomnia is important for researchers and clinicians. The objective of this study was to examine the first-night effect and the reverse first-night effect in patients with chronic primary insomnia compared with good sleeper controls. A retrospective comparison of a well-characterised sample of 50 patients with primary insomnia and 50 good sleeper controls was conducted with respect to 2 nights of polysomnography, and subjective sleep parameters in the sleep laboratory and the home setting. When comparing the first and second sleep laboratory night, a significant first-night effect was observed across both groups in the great majority of the investigated polysomnographic and subjective variables. However, patients with primary insomnia and good sleeper controls did not differ with respect to this effect. Regarding the comparison between the sleep laboratory nights and the home setting, unlike good sleeper controls, patients with primary insomnia reported an increased subjective sleep efficiency on both nights (in part due to a reduced bed time) and an increased subjective total sleep time on the second night. These results suggest that even the second sleep laboratory night does not necessarily provide clinicians and researchers with a representative insight into the sleep perception of patients with primary insomnia. Future studies should investigate whether these findings also hold for other patient populations. © 2015 European Sleep Research Society.

  16. 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)

  17. Effects of cold working ratio and stress intensity factor on intergranular stress corrosion cracking susceptibility of non-sensitized austenitic stainless steels in simulated BWR and PWR primary water

    International Nuclear Information System (INIS)

    Yaguchi, Seiji; Yonezawa, Toshio

    2012-01-01

    To evaluate the effects of cold working ratio, stress intensity factor and water chemistry on an IGSCC susceptibility of non-sensitized austenitic stainless steel, constant displacement DCB specimens were applied to SCC tests in simulated BWR and PWR primary water for the three types of austenitic stainless steels, Types 316L, 347 and 321. IGSCC was observed on the test specimens in simulated BWR and PWR primary water. The observed IGSCC was categorized into the following two types. The one is that the IGSCC observed on the same plane of the pre-fatigue crack plane, and the other is that the IGSCC observed on a plane perpendicular to the pre-fatigue crack plane. The later IGSCC fractured plane is parallel to the rolling plane of a cold rolled material. Two types of IGSCC fractured planes were changed according to the combination of the testing conditions (cold working ratio, stress intensity factor and simulated water). It seems to suggest that the most susceptible plane due to fabrication process of materials might play a significant role of IGSCC for non-sensitized cold worked austenitic stainless steels, especially, in simulated PWR primary water. Based upon evaluating on the reference crack growth rate (R-CGR) of the test specimens, the R-CGR seems to be mainly affected by cold working ratio. In case of simulated PWR primary water, it seems that the effect of metallurgical aspects dominates IGSCC susceptibility. (author)

  18. 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.

  19. Stress corrosion cracking of austenitic stainless steels in PWR primary water: an update of metallurgical investigations performed on French withdrawn components

    International Nuclear Information System (INIS)

    Boursier, J.M.; Gallet, S.; Rouillon, Y.; Bordes, P.

    2002-01-01

    Austenitic stainless steels (AISI 304, 304L, 316 and 316L) are largely used in Nuclear Power Plants because of their good resistance to corrosion and their satisfactory mechanical properties. Nevertheless, on various French PWR Nuclear Power Plants, several cases of corrosion have been encountered in auxiliary circuit portions where deleterious species and oxygen can be present. This paper focuses on the metallurgical investigations performed on pulled out components such as Canopy welds or 'dead legs' (auxiliary circuit portions connected to the main primary loops) in terms of cracking locations and degradation parameters. In addition, some comparisons between Nuclear Power Plant feedback and fundamental research and development studies are discussed, particularly in the scope of temperature, microstructure, stresses (applied and residual) and medium responsible for the degradation. (authors)

  20. 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

    International Nuclear Information System (INIS)

    Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav

    2017-01-01

    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.

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

    International Nuclear Information System (INIS)

    Martin, A.; Alvarez, D.; Cases, F.; Stelletta, S.

    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)

  2. The use of Zeolite into the controlling of Lithium concentration in the PWR primary water coolant (I) : the influences of Ca, Mg and Boric Acid concentration into the exchanges capacity of Ammonium Zeolite

    International Nuclear Information System (INIS)

    Sumijanto; Siti-Amini

    1996-01-01

    In this first part of research, the influences of calsium, magnesium and boric acid concentrations to the zeolite uptake of lithium in the PWR primary water coolant have been studied. The ammonium form of zeolite was found by modification of the natural zeolite which was originated from Bayah. The results showed that the boric acid concentration in the normal condition of PWR operation absolutely did not affects the lithium uptake. The Li uptake efficiency was influenced by the presence of Ca and Mg ions in order to the presence of cations competition which was dominated by Ca ion

  3. PWR core design calculations

    Energy Technology Data Exchange (ETDEWEB)

    Trkov, A; Ravnik, M; Zeleznik, N [Inst. Jozef Stefan, Ljubljana (Slovenia)

    1992-07-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [Slovenian] Opisali smo programski paket CORD-2, ki se uporablja pri projektnih izracunih sredice pri upravljanju tlacnovodnega reaktorja. Prikazana je uporaba paketa in racunskih postopkov za tipicne probleme, ki nastopajo pri projektiranju sredice. Primerjava glavnih rezultatov z eksperimentalnimi vrednostmi je predstavljena kot del preveritvenega procesa. (author)

  4. ROX PWR

    International Nuclear Information System (INIS)

    Akie, H.; Yamashita, T.; Shirasu, N.; Takano, H.; Anoda, Y.; Kimura, H.

    1999-01-01

    For an efficient burnup of excess plutonium from nuclear reactors spent fuels and dismantled warheads, plutonium rock-like oxide(ROX) fuel has been investigated. The ROX fuel is expected to provide high Pu transmutation capability, irradiation stability and chemical and geological stability. While, a zirconia-based ROX(Zr-ROX)-fueled PWR core has some problems of Doppler reactivity coefficient and power peaking factor. For the improvement of these characteristics, two approaches were considered: the additives such as UO 2 , ThO 2 and Er 2 O 3 , and a heterogeneous core with Zr-ROX and UO 2 assemblies. As a result, the additives UO 2 + Er 2 O 3 are found to sufficiently improve the reactivity coefficients and accident behavior, and to flatten power distribution. On the other hand, in the 1/3Zr-ROX + 2/3UO 2 heterogeneous core, further reduction of power peaking seems necessary. (author)

  5. ROX PWR

    Energy Technology Data Exchange (ETDEWEB)

    Akie, H.; Yamashita, T.; Shirasu, N.; Takano, H.; Anoda, Y.; Kimura, H. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    For an efficient burnup of excess plutonium from nuclear reactors spent fuels and dismantled warheads, plutonium rock-like oxide(ROX) fuel has been investigated. The ROX fuel is expected to provide high Pu transmutation capability, irradiation stability and chemical and geological stability. While, a zirconia-based ROX(Zr-ROX)-fueled PWR core has some problems of Doppler reactivity coefficient and power peaking factor. For the improvement of these characteristics, two approaches were considered: the additives such as UO{sub 2}, ThO{sub 2} and Er{sub 2}O{sub 3}, and a heterogeneous core with Zr-ROX and UO{sub 2} assemblies. As a result, the additives UO{sub 2}+ Er{sub 2}O{sub 3} are found to sufficiently improve the reactivity coefficients and accident behavior, and to flatten power distribution. On the other hand, in the 1/3Zr-ROX + 2/3UO{sub 2} heterogeneous core, further reduction of power peaking seems necessary. (author)

  6. PWR systems transient analysis

    International Nuclear Information System (INIS)

    Kennedy, M.F.; Peeler, G.B.; Abramson, P.B.

    1985-01-01

    Analysis of transients in pressurized water reactor (PWR) systems involves the assessment of the response of the total plant, including primary and secondary coolant systems, steam piping and turbine (possibly including the complete feedwater train), and various control and safety systems. Transient analysis is performed as part of the plant safety analysis to insure the adequacy of the reactor design and operating procedures and to verify the applicable plant emergency guidelines. Event sequences which must be examined are developed by considering possible failures or maloperations of plant components. These vary in severity (and calculational difficulty) from a series of normal operational transients, such as minor load changes, reactor trips, valve and pump malfunctions, up to the double-ended guillotine rupture of a primary reactor coolant system pipe known as a Large Break Loss of Coolant Accident (LBLOCA). The focus of this paper is the analysis of all those transients and accidents except loss of coolant accidents

  7. PWR AXIAL BURNUP PROFILE ANALYSIS

    International Nuclear Information System (INIS)

    J.M. Acaglione

    2003-01-01

    The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B andW 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001)

  8. Stress corrosion cracking growth rate of TT alloy 690 and its weld joint in simulated PWR primary water

    International Nuclear Information System (INIS)

    Yonezawa, T.

    2015-01-01

    Recently, some researchers reported that the SCC growth rate (SCCGR) of cold worked thermally treated (TT) Alloy 690 was significantly different in heat by heat. But, author has hypothesized that these high SCCGRs in cold worked TT Alloy 690 could be due to the metallurgical characteristics of these heats. In order to confirm this hypothesis, this study has been started in the author's laboratory, and the following 4 new evidences were obtained. First, microcracks of carbides and voids were observed in eutectic M 23 C 6 GB carbides (primary carbides) for cold rolled laboratory heat after as cast or lightly forged condition or for chemical composition simulated Bettis'TT Alloy 690 heat, after cold rolling, before SCC test. However, microcracks in primary carbides along grain boundaries and voids were rarely detected in the cold rolled commercial heat of TT Alloy 690 used for CRDM penetrations. Secondly, the SCCGR observed in TT Alloy 690 was different in each hot working process and each heat. Comparing the SCCGRs for all heats of cold worked TT Alloy 690, the SCCGR decreased with increasing of Vickers hardness. However, in same heats of cold worked TT Alloy 690, the SCCGR increased with increasing of Vickers hardness. Thirdly, the SCCGR in cold rolled TT Alloy 690 should be integrated by the effect of hardness or cold working ratio and by the effect of existing ratio of primary M23C6 carbides with cracks and Voids due to chemical composition and the fabrication process of TT Alloy 690. Fourthly, it is argued that the high SCCGRs in highly cold rolled TT Alloy 690 are not representative of the practical situation with TT Alloy 690 in service for CRDM adapter nozzles etc. The high SCCGR of highly cold rolled TT Alloy 690 is not thought to be an accurate tool in predicting the possibility of cracking of TT Alloy 690 for CRDM adapter nozzles. (author)

  9. Stress corrosion cracking of iron-nickel-chromium alloys in primary circuit environment of PWR-type reactors

    International Nuclear Information System (INIS)

    Boursier, Jean-Marie

    1993-01-01

    Stress corrosion cracking of Alloy 600 steam generator tubing is a great concern for pressurized water reactors. The mechanism that controls intergranular stress corrosion cracking of Alloy 600 in primary water (lithiated-borated water) has yet to be clearly identified. A study of stress corrosion cracking behaviour, which can identify the main parameters that control the cracking phenomenon, was so necessary to understand the stress corrosion cracking process. Constant extension rate tests, and constant load tests have evidenced that Alloy 600 stress corrosion cracking involves firstly an initiation period, then a slow propagation stage with crack less than 50 to 80 micrometers, and finally a rapid propagation stage leading to failure. The influence of mechanical parameters have shown the next points: - superficial strain hardening and cold work have a strong effect of stress corrosion cracking resistance (decrease of initiation time and increase of crack growth rate), - strain rate was the most suitable parameter for describing the different stage of propagation. The creep behaviour of alloy 600 has shown an increase of creep rate in primary water compared to air, which implies a local interaction plasticity/corrosion. An assessment of the durations of the initiation and the propagation stages was attempted for the whole uniaxial tensile tests, using the macroscopic strain rate: - the initiation time is less than 100 hours and seems to be an electrochemical process, - the durations of the propagation stage are strongly dependent on the strain rate. The behaviour in high primary water temperature of Alloys 690 and 800, which replace Alloy 600, was studied to appraise their margin, and validate their choice. Then the last chapter has to objective to evaluate the crack tip strain rate, in order to better describe the evolution of the different stages of cracking. (author) [fr

  10. Vibration monitoring of the primary piping systems during the hot functional tests of the Mulheim-Karlich PWR

    International Nuclear Information System (INIS)

    Bauernfeind, V.; Bloem, T.; Pache, W.; Diederich, H.J.

    1989-01-01

    During the hot functional tests of the Muelheim--Kaerlich first-of-a-kind plant, vibration measurements were made on the reactor pressure vessel and its' internals and on the primary piping system and main coolant pumps. This paper contains results of the measurements taken on the pipes and the pumps with an interpretation of these measurements based on an analytical model of the primary system. The main aim of the measurement program is to confirm that the components, which are of new design, are adequately dimensioned for the operational vibration loads during the service life of the reactor. In addition, the vibrational modes of the hot lines, the steam generators and the pumps with the adjacent cold lines were determined. These values were compared with the analytically calculated resonance frequencies and eigenforms. Good agreement was found. In the course of these comparisons, information on the modelling of the supporting structures and the efficiency of the damping elements during normal operation was obtained

  11. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 3: nonseismic stress analysis. Final report

    International Nuclear Information System (INIS)

    Chan, A.L.; Curtis, D.J.; Rybicki, E.F.; Lu, S.C.

    1981-08-01

    This volume describes the analyses used to evaluate stresses due to loads other than seismic excitations in the primary coolant loop piping of a selected four-loop pressurized water reactor nuclear power station. The results of the analyses are used as input to a simulation procedure for predicting the probability of pipe fracture in the primary coolant system. Sources of stresses considered in the analyses are pressure, dead weight, thermal expansion, thermal gradients through the pipe wall, residual welding, and mechanical vibrations. Pressure and thermal transients arising from plant operations are best estimates and are based on actual plant operation records supplemented by specified plant design conditions. Stresses due to dead weight and thermal expansion are computed from a three-dimensional finite element model that uses a combination of pipe, truss, and beam elements to represent the reactor coolant loop piping, reactor pressure vessel, reactor coolant pumps, steam generators, and the pressurizer. Stresses due to pressure and thermal gradients are obtained by closed-form solutions. Calculations of residual stresses account for the actual heat impact, welding speed, weld preparation geometry, and pre- and post-heat treatments. Vibrational stresses due to pump operation are estimated by a dynamic analysis using existing measurements of pump vibrations

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

    International Nuclear Information System (INIS)

    Fields, S.R.

    1979-06-01

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

  13. Vibration monitoring of the primary piping system during the hot functional tests of the Muelheim-Kaerlich PWR

    International Nuclear Information System (INIS)

    Bauernfeind, V.; Bloem, T.; Pache, W.; Diederich, H.J.

    1992-01-01

    During the hot functional tests of the Muelheim-Kaerlich plant, which was the first plant of its type, vibration measurements were made on the reactor pressure vessel and its internal parts and on the primary piping system and the main coolant pumps. This paper contains the results of the measurements taken on the pipes and the pumps with an interpretation of these measurements based on an analytical model of the primary system. The main aim of the measurement programs is to confirm that the components, which are of new structural design, are adequately dimensioned for the operational vibration loads during the service life of the reactor. In addition, the vibrational modes of the hot lines, the steam generators and the pumps with the adjacent cold lines were determined. These values were compared with the analytically calculated resonance frequencies and eigenforms. A good correspondence was found. In the course of these comparisons, information about the modelling of the supporting structures and the efficiency of the damping elements during normal operation was obtained. The vibration of the main coolant pumps was also continuously monitored. The pump surveillance system for each pump includes two non-contacting displacement sensors for measuring the kinetic shaft orbit, as well as velocity sensors for recording the vibrational velocity of the pump motor housing. During the continuous monitoring, it was checked whether the signal amplitudes remained within the allowable limits. In addition the frequency content of the signals was determined periodically. In this way deviations could be detected immediately and be explained by means of subsequent correlation analysis. Thus amplitude changes resulting from resonance effects were identified. (orig.)

  14. PWR degraded core analysis

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1982-04-01

    A review is presented of the various phenomena involved in degraded core accidents and the ensuing transport of fission products from the fuel to the primary circuit and the containment. The dominant accident sequences found in the PWR risk studies published to date are briefly described. Then chapters deal with the following topics: the condition and behaviour of water reactor fuel during normal operation and at the commencement of degraded core accidents; the generation of hydrogen from the Zircaloy-steam and the steel-steam reactions; the way in which the core deforms and finally melts following loss of coolant; debris relocation analysis; containment integrity; fission product behaviour during a degraded core accident. (U.K.)

  15. Maturity of the PWR

    International Nuclear Information System (INIS)

    Bacher, P.; Rapin, M.; Aboudarham, L.; Bitsch, D.

    1983-03-01

    Figures illustrating the predominant position of the PWR system are presented. The question is whether on the basis of these figures the PWR can be considered to have reached maturity. The following analysis, based on the French program experience, is an attempt to pinpoint those areas in which industrial maturity of the PWR has been attained, and in which areas a certain evolution can still be expected to take place

  16. Estimation of maximum pressure in small containments of PWR reactors due to loss of coolant accident in primary circuit; Estimativa da pressao maxima em contencoes de reatores PWR de pequeno porte devido a um acidente de perda de refrigerante no circuito primario

    Energy Technology Data Exchange (ETDEWEB)

    Mendes Neto, Teofilo [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil); Moreira, Joao Manoel Losada [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)

    2000-07-01

    This work studies the problem of containment pressurization after a LOCA in reactors with small containment free volumes. The relationship between the reactor power and the containment free volume is described with the ratio between the volumes of the primary circuit and of the containment. The maximum pressure in a containment, following a LOCA, obtained after a correlation based on large containment PWR, is around 185 psia for a primary circuit and containment volumes ratio of 0.025. For the same problem, calculations with the CONTEMPT-LT code produced a maximum pressure of 162 psia. The behavior of the temperature after a LOCA to the containment, as a function of the ratio between the primary circuit and containment volume, is such that it increases reaching asymptotically to a maximum; differently, the pressure increases almost linearly with the ratio of volumes. (author)

  17. Comprehensive exergetic and economic comparison of PWR and hybrid fossil fuel-PWR power plants

    International Nuclear Information System (INIS)

    Sayyaadi, Hoseyn; Sabzaligol, Tooraj

    2010-01-01

    A typical 1000 MW Pressurized Water Reactor (PWR) nuclear power plant and two similar hybrid 1000 MW PWR plants operate with natural gas and coal fired fossil fuel superheater-economizers (Hybrid PWR-Fossil fuel plants) are compared exergetically and economically. Comparison is performed based on energetic and economic features of three systems. In order to compare system at their optimum operating point, three workable base case systems including the conventional PWR, and gas and coal fired hybrid PWR-Fossil fuel power plants considered and optimized in exergetic and exergoeconomic optimization scenarios, separately. The thermodynamic modeling of three systems is performed based on energy and exergy analyses, while an economic model is developed according to the exergoeconomic analysis and Total Revenue Requirement (TRR) method. The objective functions based on exergetic and exergoeconomic analyses are developed. The exergetic and exergoeconomic optimizations are performed using the Genetic Algorithm (GA). Energetic and economic features of exergetic and exergoeconomic optimized conventional PWR and gas and coal fired Hybrid PWR-Fossil fuel power plants are compared and discussed comprehensively.

  18. Study on a corrosion products behavior with zinc injection in PWR primary circuit based on the crud investigation in actual plant

    International Nuclear Information System (INIS)

    Hisamune Kenji; Nambu Tohru; Akutagawa Daisuke; Nagata Nobuaki; Shimizu, Yuichi; Nagamine, Kunitaka; Kogawa Noritaka

    2012-09-01

    Zinc injection into the reactor coolant of Pressurized Water Reactor (PWR) is one of the most effective techniques for removing radioactive cobalt from the chromite type spinel oxides in the inner layer of reactor coolant component surfaces. Many documents have reported that zinc injection applied plants have generally achieved positive dose rate reduction effects for the reactor coolant components and piping. Tsuruga Nuclear Power Plant unit 2 (hereafter Tsuruga unit 2) which applied zinc injection in 2005 has also achieved good dose reduction effects in most primary coolant piping. However, some piping without dose-rate reduction effects exist, it has come to an issue to plan radiation exposure reduction measures for the works around relevant piping. To plan a reasonable dose-rate reduction measures for respective components and piping in primary systems, it is considered necessary to understand the dose-rate reduction effects of zinc injection mechanically. Therefore, the surface oxides/deposits on piping and fuel cladding tubes with operational histories under zinc injection were investigated. As a result of the investigation, the changes in specific aspects of oxide layers and CRUDs are found as follow: - Owing to long-term zinc injection, deposits and oxides on the primary coolant piping are extremely reduced, the boundary between the outer oxide layer and the inner oxide layer become unclear with the thinning of the outer oxide layers. - At a Pressurizer Spray Piping, the dose-rate increases than before zinc injection. The oxide layers have the same 3-layer construction as before zinc injection, and the surface concentration of the radioactive cobalt is high. - At the fuel cladding tube surfaces, what is observed is the amount of metal nickel decrease and the chromium composition increase in the spinel type oxide From these results, it is suggested that, the oxide layers composed principally of zinc chromite spinel grow up on the pipe surface with zinc

  19. 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.).

  20. The PWR cores management

    International Nuclear Information System (INIS)

    Barral, J.C.; Rippert, D.; Johner, J.

    2000-01-01

    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.)

  1. Improvement of PWR reliability by corrosion prevention

    International Nuclear Information System (INIS)

    Takamatsu, Hiroshi

    1999-01-01

    Since first PWR in Japan started commercial operation in 1970, we have encountered the various modes of corrosion on primary and secondary side components. We have paid much efforts for resolving these corrosion problems, that is, investigating the causes of corrosion and establishing the countermeasures for these corrosion. We summarize these efforts in this article. (author)

  2. Evaluation of ultimate load bearing capacity of the primary containment of a typical 540 MWe Indian PHWR

    International Nuclear Information System (INIS)

    Ray, Indrajit; Roy, Raghupati; Verma, U.S.P.; Warudkar, A.S.

    2003-01-01

    This paper presents the analysis of the Inner Containment Structure (ICS) of a typical 540 MWe Indian PHWR for the purpose of evaluating its ultimate load bearing capacity (ULBC) under beyond postulated design basis accident (DBA) scenario. The methodology adopted for the non-linear analysis of the prestressed concrete ICS including the various issues, viz. behaviour of concrete under compression and tension, tension stiffening, cracked shear modulus etc. have also been discussed in this paper. The effect of accident temperature on ULBC has been studied and discussed in this paper. This paper also discusses about the study carried out for mesh sensitivity of the finite element (FE) discretization on ULBC of ICS in the non-linear range. Based on the detailed analysis, the factor of safety of the ICS under beyond postulated DBA scenario has been evaluated. (author)

  3. Evaluation of ultimate load bearing capacity of the primary containment of a typical first generation 220 MWe Indian PHWRs

    International Nuclear Information System (INIS)

    Singh, A.K.; Ray, I.; Roy, R.; Garg, R.P.; Verma, U.S.P.

    2005-01-01

    This paper presents the analysis of the Inner Containment Structure (ICS) of a typical first generation 220 MWe Indian PHWR for the purpose of evaluating its ultimate load bearing capacity (ULBC) beyond postulated design basis accident (DBA) scenario. The first generation ICS of Indian PHWRs are made of cylindrical wall capped with prestressed/reinforced cellular containment slab, which is connected monolithically to outer containment (OC) Wall to provide clamping effect during postulated DBA scenario. This paper discusses the simulation of construction sequence in analytical model, which is a very important aspect from point of view of capturing the residual stresses generated in the structure. The methodology adopted for the non-linear analysis of the prestressed concrete ICS including the various issues, viz. behaviour of concrete under compression and tension, tension stiffening, cracked shear modulus etc. have also been discussed in this paper. The effect of accident temperature on ULBC has been studied and discussed in this paper. This paper also discusses the study of mesh sensitivity of the finite element (FE) discretisation on ULBC of ICS in the non-linear range. Based on the detailed analysis, the factor of safety of the ICS beyond postulated DBA scenario has been evaluated. (authors)

  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. Organization patterns of PWR power plants

    International Nuclear Information System (INIS)

    Leicman, J.

    1980-01-01

    Organization patterns are shown for the St. Lucia 1, North Anna, Sequoyah, and Beaver Valley nuclear power plants, for a typical PWR power plant in the USA, for the Biblis/RWE-KWU nuclear power plants and for a four-unit nuclear power plant operated by Electricite de France as well as for the Loviisa power plant. Organization patterns are also shown for relatively independent and non-independent nuclear power plants according to IAEA recommendations. (J.P.)

  6. The vocabulary profile of Slovak children with primary language impairment compared to typically developing Slovak children measured by LITMUS-CLT.

    Science.gov (United States)

    Kapalková, Svetlana; Slančová, Daniela

    2017-01-01

    This study compared a sample of children with primary language impairment (PLI) and typically developing age-matched children using the crosslinguistic lexical tasks (CLT-SK). We also compared the PLI children with typically developing language-matched younger children who were matched on the basis of receptive vocabulary. Overall, statistical testing showed that the vocabulary of the PLI children was significantly different from the vocabulary of the age-matched children, but not statistically different from the younger children who were matched on the basis of their receptive vocabulary size. Qualitative analysis of the correct answers revealed that the PLI children showed higher rigidity compared to the younger language-matched children who are able to use more synonyms or derivations across word class in naming tasks. Similarly, an examination of the children's naming errors indicated that the language-matched children exhibited more semantic errors, whereas PLI children showed more associative errors.

  7. Simulation model for the dynamic behavior of the hydraUlic circuito of PWR reactors

    International Nuclear Information System (INIS)

    Hirdes, V.R.T.R.

    1987-01-01

    The present work consist of the development of a computer code for the simulations of hydraulic transients caused by stoppages of the primary coolant pumps of nuclear reactors and it applied to the hydraulic circuits typical of PWR reactor. The code calculates the time-histories of the mass flux, rotation speed, electric and hydraulic torque and dynamic head of the pumps. It can be used for any combination of active and inactive pumps. Several transients were analysed and the results were compared with comparared with data from the Angra-I nuclear power plant. The results were considered satisfactory. (author) [pt

  8. Next generation PWR

    International Nuclear Information System (INIS)

    Tanaka, Toshihiko; Fukuda, Toshihiko; Usui, Shuji

    2001-01-01

    Development of LWR for power generation in Japan has been intended to upgrade its reliability, safety, operability, maintenance and economy as well as to increase its capacity in order, since nuclear power generation for commercial use was begun on 1970, to steadily increase its generation power. And, in Japan, ABWR (advanced BWR) of the most promising LWR in the world, was already used actually and APWR (advanced PWR) with the largest output in the world is also at a step of its actual use. And, development of the APWR in Japan was begun on 1980s, and is at a step of plan on construction of its first machine at early of this century. However, by large change of social affairs, economy of nuclear power generation is extremely required, to be positioned at an APWR improved development reactor promoted by collaboration of five PWR generation companies and the Mitsubishi Electric Co., Ltd. Therefore, on its development, investigation on effect of change in social affairs on nuclear power stations was at first carried out, to establish a design requirement for the next generation PWR. Here were described on outline, reactor core design, safety concept, and safety evaluation of APWR+ and development of an innovative PWR. (G.K.)

  9. Analysis of breaks in pipe connections to the hot leg and to the loop seal in the primary system of Ringhals 2 PWR

    International Nuclear Information System (INIS)

    Nilsson, L.; Sjoeberg, A.

    1987-01-01

    Analysis has been made of seven different cases of breaks in pipes connected to the hot leg and to the loop seal in Ringhals 2 PWR. The pipes, in which guillotine breaks are postulated, have nominal diameters ranging from 1 to 14 inches. The purpose of the analysis is to supplement the basis for a review of the inspection procedures for the safety of pressure vessels prescribed by SKI. A similar analysis already exists concerning breaks in the cold leg connections. The analysis has been made using the thermal hydraulic computer code RELAPS/MOD2 and with best estimate assumptions. This means that normal operating conditions have been adopted for the input to the calculations. However, the capacity of the safety injection system was assumed to be reduced by having one pump not operating each of the low pressure and high pressure safety injection system. The results of the analysis are presented in tables and as computer plots. The analysis shows that the consequences with respect to increased fuel rod and cladding temperatures are quite harmless. Only the two cases with the largest break sizes lead to critical heat flux (CHF) and increased temperatures for the hottest rods in the core. The peak cladding temperature was 636 degrees C for the 12 inch break. In both cases rewetting occurred within 25 s of accident initiation. In the cases with breaks in connections of 6 inch diameter and smaller the RELAP5 calculations indicated a substantial margin to CHF throughout the transient. (authors)

  10. PWR and WWER fuel performance. A comparison of major characteristics

    International Nuclear Information System (INIS)

    Weidinger, H.

    2006-01-01

    PWR and WWER fuel technologies have the same basic performance targets: most effective use of the energy stored in the fuel and highest possible reliability. Both fuel technologies use basically the same strategies to reach these targets: 1) Optimized reload strategies; 2) Maximal use of structural material with low neutron cross sections; 3) Decrease the fuel failure frequency towards a 'zero failure' performance by understanding and eliminating the root causes of those defects. The key driving force of the technology of both, PWR and WWER fuel is high burn-up. Presently a range of 45 - 50 MWD/kgU have been reached commercially for PWR and WWER fuel. The main technical limitations to reach high burn-up are typically different for PWR and WWER fuel: for PWR fuel it is the corrosion and hydrogen uptake of the Zr-based materials; for WWER fuel it is the mechanical and dimensional stability of the FA (and the whole core). Corrosion and hydrogen uptake of Zr-materials is a 'non-problem' for WWER fuel. Other performance criteria that are important for high burn-up are the creep and growth behaviour of the Zr materials and the fission gas release in the fuel rod. There exists a good and broad data base to model and design both fuel types. FA and fuel rod vibration appears to be a generic problem for both fuel types but with more evidence for PWR fuel performance reliability. Grid-to-rod fretting is still a major issue in the fuel failure statistics of PWR fuel. Fuel rod cladding defects by debris fretting is no longer a key problem for PWR fuel, while it still appears to be a significant root cause for WWER fuel failures. 'Zero defect' fuel performance is achievable with a high probability, as statistics for US PWR and WWER-1000 fuel has shown

  11. Minor actinide transmutation on PWR burnable poison rods

    International Nuclear Information System (INIS)

    Hu, Wenchao; Liu, Bin; Ouyang, Xiaoping; Tu, Jing; Liu, Fang; Huang, Liming; Fu, Juan; Meng, Haiyan

    2015-01-01

    Highlights: • Key issues associated with MA transmutation are the appropriate loading pattern. • Commercial PWRs are the only choice to transmute MAs in large scale currently. • Considerable amount of MA can be loaded to PWR without disturbing k eff markedly. • Loading MA to PWR burnable poison rods for transmutation is an optimal loading pattern. - Abstract: Minor actinides are the primary contributors to long term radiotoxicity in spent fuel. The majority of commercial reactors in operation in the world are PWRs, so to study the minor actinide transmutation characteristics in the PWRs and ultimately realize the successful minor actinide transmutation in PWRs are crucial problem in the area of the nuclear waste disposal. The key issues associated with the minor actinide transmutation are the appropriate loading patterns when introducing minor actinides to the PWR core. We study two different minor actinide transmutation materials loading patterns on the PWR burnable poison rods, one is to coat a thin layer of minor actinide in the water gap between the zircaloy cladding and the stainless steel which is filled with water, another one is that minor actinides substitute for burnable poison directly within burnable poison rods. Simulation calculation indicates that the two loading patterns can load approximately equivalent to 5–6 PWR annual minor actinide yields without disturbing the PWR k eff markedly. The PWR k eff can return criticality again by slightly reducing the boric acid concentration in the coolant of PWR or removing some burnable poison rods without coating the minor actinide transmutation materials from PWR core. In other words, loading minor actinide transmutation material to PWR does not consume extra neutron, minor actinide just consumes the neutrons which absorbed by the removed control poisons. Both minor actinide loading patterns are technically feasible; most importantly do not need to modify the configuration of the PWR core and

  12. PWR Users Group 10 CFR 61 Waste Form Requirements Compliance Test Program

    International Nuclear Information System (INIS)

    Rosenlof, R.C.

    1985-01-01

    In January of 1984, a PWR Users Group was formed to initiate a 10 CFR 61 Waste Form Requirements Compliance Test Program on a shared cost basis. The original Radwaste Solidification Systems sold by ATCOR ENGINEERED SYSTEMS, INC. to the utilities were required to produce a free-standing monolith with no free water. None of the other requirements of 10 CFR 61 had to be met. Current regulations, however, have substantially expanded the scope of the waste form acceptance criteria. These new criteria required that generators of radioactive waste demonstrate the ability to produce waste forms which meet certain chemical and physical requirements. This paper will present the test program used and the results obtained to insure 10 CFR 61 compliance of the three (3) typical waste streams generated by the ATCOR PWR Users Group's plants. The primary objective of the PWR Users Group was not to maximize waste loading within the masonry cement solidification media, but to insure that the users Radwaste Solidification System is capable of producing waste forms which meet the waste form criteria of 10 CFR 61. A description of the laboratory small sample certification program and the actual full scale pilot plant verification approach used is included in this paper. Also included is a discussion of the development of a Process Control Program to ensure the reproducibility of the test results with actual waste

  13. Typical entanglement

    Science.gov (United States)

    Deelan Cunden, Fabio; Facchi, Paolo; Florio, Giuseppe; Pascazio, Saverio

    2013-05-01

    Let a pure state | ψ> be chosen randomly in an NM-dimensional Hilbert space, and consider the reduced density matrix ρ A of an N-dimensional subsystem. The bipartite entanglement properties of | ψ> are encoded in the spectrum of ρ A . By means of a saddle point method and using a "Coulomb gas" model for the eigenvalues, we obtain the typical spectrum of reduced density matrices. We consider the cases of an unbiased ensemble of pure states and of a fixed value of the purity. We finally obtain the eigenvalue distribution by using a statistical mechanics approach based on the introduction of a partition function.

  14. The PWR spectral code GELS. Pt. 1

    International Nuclear Information System (INIS)

    Penndorf, K.; Schult, F.; Schulz, G.

    1976-01-01

    The code procedures group constant libraries for the static PWR design of whatever fuel cycle - Uranium, Thorium, or Plutonium. The whole reach of temperatures is covered and the treatment of strong lumped absorbers as control or burnable poison pins is included. The main features are: 1) Good accuracy in spite of not fitting the material data to critical experiments; 2) speed and relatively low computer equipment; 3) restriction to PWR's only. In case of demands for higher accuracy there is a further restriction concerning the library data of the epithermal resonance absorbers: They are strictly valid only for several special lattice geometrics. Three samples are given each representing a typical application of the code. Two of them likewise are demonstrations of recalculated experiments. (orig.) [de

  15. Fuel management optimization for a PWR

    International Nuclear Information System (INIS)

    Dumas, M.; Robeau, D.

    1981-04-01

    This study is aimed to optimize the refueling pattern of a PWR. Two methods are developed, they are based on a linearized form of the optimization problem. The first method determines a feasible solution in two steps; in the first one the original problem is replaced by a relaxed one which is solved by the Method of Approximation Programming. The second step is based on the Branch and Bound method to find the feasible solution closest to the solution obtained in the first step. The second method starts from a given refueling pattern and tries to improve this pattern by the calculation of the effects of 2 by 2, 3 by 3 and 4 by 4 permutations on the objective function. Numerical results are given for a typical PWR refueling using the two methods

  16. 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)

  17. Scaling studies - PWR

    International Nuclear Information System (INIS)

    Sonneck, G.

    1983-05-01

    A RELAP 4/MOD 6 study was made based on the blowdown phase of the intermediate break experiment LOFT L5-1. The method was to set up a base model and to vary parametrically some areas where it is known or suspected that LOFT differs from a commercial PWR. The aim was not to simulate LOFT or a PWR exactly but to understand the influence of the following parameters on the thermohydraulic behaviour of the system and the clad temperature: stored heat in the downcomer (LOFT has rather large filler blocks in this part of the pressure vessel); bypass between downcomer and upper plenum; and core length. The results show that LOFT is prototypical for all calculated blowdowns. As the clad temperatures decrease with decreasing stored energy in the downcomer, increased bypass and increased core length, LOFT results seem to be realistic as long as realistic bypass sizes are considered; they are conservative in the two other areas. (author)

  18. Plutonium recycling in PWR

    International Nuclear Information System (INIS)

    Youinou, G.; Girieud, R.; Guigon, B.

    2000-01-01

    Two concepts of 100% MOX PWR cores are presented. They are designed such as to minimize the consequences of the introduction of Pu on the core control. The first one has a high moderation ratio and the second one utilizes an enriched uranium support. The important design parameters as well as their capabilities to multi recycle Pu are discussed. We conclude with the potential interest of the two concepts. (author)

  19. The integrated PWR

    International Nuclear Information System (INIS)

    Gautier, G.M.

    2002-01-01

    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.)

  20. Reactor control system. PWR

    International Nuclear Information System (INIS)

    2009-01-01

    At present, 23 units of PWR type reactors have been operated in Japan since the start of Mihama Unit 1 operation in 1970 and various improvements have been made to upgrade operability of power stations as well as reliability and safety of power plants. As the share of nuclear power increases, further improvements of operating performance such as load following capability will be requested for power stations with more reliable and safer operation. This article outlined the reactor control system of PWR type reactors and described the control performance of power plants realized with those systems. The PWR control system is characterized that the turbine power is automatic or manually controlled with request of the electric power system and then the nuclear power is followingly controlled with the change of core reactivity. The system mainly consists of reactor automatic control system (control rod control system), pressurizer pressure control system, pressurizer water level control system, steam generator water level control system and turbine bypass control system. (T. Tanaka)

  1. AGR v PWR

    International Nuclear Information System (INIS)

    Green, D.

    1986-01-01

    When the Central Electricity Generating Board (CEGB) invited tenders and placed a contract for the Advanced Gas Cooled Reactor (AGR) at Dungeness B in 1965 -preferring it to the Pressurised Water Reactor (PWR) -the AGR was lamentably ill developed. The effects of the decision were widely felt, for it took the British nuclear industry off the light water reactor highway of world reactor business and up and idiosyncratic private highway of its own, excluding it altogether from any material export business in the two decades which followed. Yet although the UK may have made wrong decisions in rejecting the PWR in 1965, that does not mean that it can necessarily now either correct them, or redeem their consequence, by reversing the choice in 1985. In the 20 years since 1965 the whole world economic and energy picture has been transformed and the national picture with it. Picking up the PWR now could prove as big a disaster as rejecting it may have been in 1965. (author)

  2. Water chemistry in PWR

    International Nuclear Information System (INIS)

    Abe, Kenji

    1987-01-01

    This article outlines major features and basic concept of the secondary system of PWR's and water properties control measures adopted in recent PWR plants. The secondary system of a PWR consists of a condenser cooling pipe (aluminum-brass, titanium, or stainless steel), low-pressure make-up water heating pipe (aluminum-brass or stainless steel), high-ressure make-up water heating pipe (cupro-nickel or stainless steel), steam generator heat-transfer pipe (Inconel 600 or 690), and bleed/drain pipe (carbon steel, low alloy steel or stainless steel). Other major pipes and equipment are made of carbon steel or stainless steel. Major troubles likely to be caused by water in the secondary system include reduction in wall thickness of the heat-transfer pipe, stress corrosion cracking in the heat-transfer pipe, and denting. All of these are caused by local corrosion due to concentration of purities contained in water. For controlling the water properties in the secondary system, it is necessary to prevent impurities from entering the system, to remove impurities and corrosion products from the system, and to prevent corrosion of apparatus making up the system. Measures widely adopted for controlling the formation of IGA include the addition of boric acid for decreasing the concentration of free alkali and high hydrazine operation for providing a highly reducing atmospere. (Nogami, K.)

  3. Implementation in free software of the PWR type university nucleo electric simulator (SU-PWR)

    International Nuclear Information System (INIS)

    Valle H, J.; Hidago H, F.; Morales S, J.B.

    2007-01-01

    Presently work is shown like was carried out the implementation of the University Simulator of Nucleo-electric type PWR (SU-PWR). The implementation of the simulator was carried out in a free software simulation platform, as it is Scilab, what offers big advantages that go from the free use and without cost of the product, until the codes modification so much of the system like of the program with the purpose of to improve it or to adapt it to future routines and/or more advanced graphic interfaces. The SU-PWR shows the general behavior of a PWR nuclear plant (Pressurized Water Reactor) describing the dynamics of the plant from the generation process of thermal energy in the nuclear fuel, going by the process of energy transport toward the coolant of the primary circuit the one which in turn transfers this energy to the vapor generators of the secondary circuit where the vapor is expanded by means of turbines that in turn move the electric generator producing in this way the electricity. The pressurizer that is indispensable for the process is also modeled. Each one of these stages were implemented in scicos that is the Scilab tool specialized in the simulation. The simulation was carried out by means of modules that contain the differential equation that mathematically models each stage or equipment of the PWR plant. The result is a series of modules that based on certain entrances and characteristic of the system they generate exits that in turn are the entrance to other module. Because the SU-PWR is an experimental project in early phase, it is even work and modifications to carry out, for what the models that are presented in this work can vary a little the being integrated to the whole system to simulate, but however they already show clearly the operation and the conformation of the plant. (Author)

  4. PWR: 10 years after and perspectives

    International Nuclear Information System (INIS)

    1990-01-01

    These proceedings of the SFEN days on PWR (Ten years after and perspectives) comprise 13 conferences bearing on: - From the occurential approach to the state approach - Evolution of calculating tools - Human factors and safety - Reactor safety in the PWR 2000 - The PWR and the electrical power grid load follow - Fuel aspect of PWR management - PWR chemistry evolution - Balance of radiation protection - PWR modifications balance and influence on reactor operation - Design and maintenance of reactor components: 4 conferences [fr

  5. Cylindrization of a PWR core for neutronic calculations

    International Nuclear Information System (INIS)

    Santos, Rubens Souza dos

    2005-01-01

    In this work we propose a core cylindrization, starting from a PWR core configuration, through the use of an algorithm that becomes the process automated in the program, independent of the discretization. This approach overcomes the problem stemmed from the use of the neutron transport theory on the core boundary, in addition with the singularities associated with the presence of corners on the outer fuel element core of, existents in the light water reactors (LWR). The algorithm was implemented in a computational program used to identification of the control rod drop accident in a typical PWR core. The results showed that the algorithm presented consistent results comparing with an production code, for a problem with uniform properties. In our conclusions, we suggest, for future works, for analyzing the effect on mesh sizes for the Cylindrical geometry, and to compare the transport theory calculations versus diffusion theory, for the boundary conditions with corners, for typical PWR cores. (author)

  6. Best-estimate analysis of a loss-of-coolant accident in a four-loop US PWR using TRAC-PD2

    International Nuclear Information System (INIS)

    Ireland, J.R.

    1982-01-01

    A 200-percent double-ended cold-leg break loss-of-coolant accident (LOCA) in a typical US pressurized water reactor (PWR) was simulated using the Transient Reactor Analysis Code (TRAC-PD2). The reactor system modeled represented a typical US PWR with four loops (three intact, one broken) and cold-leg emergency-core-cooling systems (ECCS). The finely noded TRAC model employed 440 three dimensional (r, THETA, z) vessel cells along with approximately 300 one-dimensional cells that modeled the primary system loops. The calculated peak-clad temperature of 950 0 K occurred during blowdown and the clad temperature excursion was terminated at 175 s, when complete core quenching occurred. Accumulator flows were initiated at 10 s, when the system pressure reached 4.08 MPa, and the refill phase ended at 36 s when the lower plenum refilled. During reflood, both bottom and falling film quench fronts were calculated

  7. Materials performance in operating PWR steam generators

    International Nuclear Information System (INIS)

    Weeks, J.R.

    1975-01-01

    The Inconel-600 tubing in operating PWR steam generators has developed leaks due to intergranular stress corrosion cracking or a general wastage attack, originating from the secondary side of the tubing. Corrosion has been limited to those areas of the steam generators where limited coolant circulation and high heat flux have caused impurities to concentrate. Wastage or pitting attack has always been associated with local concentration of sodium hydrogen phosphates, whereas stress corrosion has been associated with local concentration of sodium or potassium hydroxides. The only instance of stress corrosion originating from the primary side occurred on cold-worked tubing when hydrogen was not added to getter oxygen, and LiOH was not added to raise the pH of the primary coolant. All PWR manufacturers are now recommending that the phosphate treatment of the secondary coolant be abandoned in favor of an all-volatile treatment. Experience in operating plants has shown, however, that removal of phosphate-rich sludge deposits is difficult, and that further wastage and/or intergranular stress corrosion may develop; the residual sodium phosphates gradually convert by reaction with corrosion product hydroxides to sodium hydroxide, which remains concentrated in the limited flow areas. Improvements in circulation patterns have been achieved by inserting flow baffles in some PWR steam generators. Inservice monitoring by eddy current techniques is useful for detecting corrosion-induced defects in the tubing, but irreproducibility in field examinations can lead to uncertainties interpreting the results. (U.S.)

  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. Thermal-hydraulic study of integrated steam generator in PWR

    International Nuclear Information System (INIS)

    Osakabe, Masahiro

    1989-01-01

    One of the safety aspects of innovative reactor concepts is the integration of steam generators (SGs) into the reactor vessel in the case of the pressurized water reactor (PWR). All of the reactor system components including the pressurizer are within the reactor vessel in the SG integrated PWR. The simple heat transfer code was developed for the parametric study of the integrated SG. The code was compared to the once-through 19-tube SG experiment and the good agreement between the experimental results and the code predictions was obtained. The assessed code was used for the parametric study of the integrated once-through 16 m-straight-tube SG installed in the annular downcomer. The proposed integrated SG as a first attempt has approximately the same tube size and pitch as the present PWR and the SG primary and secondary sides in the present PWR is inverted in the integrated PWR. Based on the study, the reactor vessel size of the SG integrated PWR was calculated. (author)

  10. French PWR safety philosophy

    International Nuclear Information System (INIS)

    Conte, M.

    1986-05-01

    Increasing knowledge and lessons learned from starting and operating experience of French nuclear power plants, completed by the experience learned from the operation of foreign reactors, has contributed to the improvement of French PWR design and safety philosophy. Based on a deterministic approach, the French safety analysis was progressively completed by a probabilistic approach, each of them having possibilities and limits. As a consequence of the global risk objective set in 1977 for nuclear reactors, safety analysis was extended to the evaluation of events more complex than the conventional ones, and later to the evaluation of the feasibility of the offsite emergency plans in case of severe accidents

  11. 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.

  12. PWR decontamination feasibility study

    International Nuclear Information System (INIS)

    Silliman, P.L.

    1978-01-01

    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

  13. Modelling activity transport behavior in PWR plant

    International Nuclear Information System (INIS)

    Henshaw, Jim; McGurk, John; Dickinson, Shirley; Burrows, Robert; Hinds, Kelvin; Hussey, Dennis; Deshon, Jeff; Barrios Figueras, Joan Pau; Maldonado Sanchez, Santiago; Fernandez Lillo, Enrique; Garbett, Keith

    2012-09-01

    The activation and transport of corrosion products around a PWR circuit is a major concern to PWR plant operators as these may give rise to high personnel doses. The understanding of what controls dose rates on ex-core surfaces and shutdown releases has improved over the years but still several questions remain unanswered. For example the relative importance of particle and soluble deposition in the core to activity levels in the plant is not clear. Wide plant to plant and cycle to cycle variations are noted with no apparent explanations why such variations are observed. Over the past few years this group have been developing models to simulate corrosion product transport around a PWR circuit. These models form the basis for the latest version of the BOA code and simulate the movement of Fe and Ni around the primary circuit. Part of this development is to include the activation and subsequent transport of radioactive species around the circuit and this paper describes some initial modelling work in this area. A simple model of activation, release and deposition is described and then applied to explain the plant behaviour at Sizewell B and Vandellos II. This model accounts for activation in the core, soluble and particulate activity movement around the circuit and for activity capture ex-core on both the inner and outer oxides. The model gives a reasonable comparison with plant observations and highlights what controls activity transport in these plants and importantly what factors can be ignored. (authors)

  14. Safety considerations of PWR's

    International Nuclear Information System (INIS)

    Arnold, W.H. Jr.

    1977-01-01

    The safety of the central station pressurized water reactor is well established and substantiated by its excellent operating record. Operating data from 55 reactors of this type have established a record of safe operating history unparalleled by any modern large scale industry. The 186 plants under construction require a continuing commitment to maintain this outstanding record. The safety of the PWR has been further verified by the recently completed Reactor Safety Study (''Rasmussen'' Report). Not only has this study confirmed the exceptionally low risk associated with PWR operation, it has also introduced a valuable new tool in the decision making process. PWR designs, utilizing the philosophy of defense in depth, provide the bases for evaluating margins of safety. The design of the reactor coolant system, the containment system, emergency core cooling system and other related systems and components provide substantial margins of safety under both normal and postulated accident conditions even considering simultaneous effects of earthquakes and other environmental phenomena. Margins of safety in the assessment of various postulated accident conditions, with emphasis on the postulated loss of reactor coolant accident (LOCA), have been evaluated in depth as exemplified by the comprehensive ECCS rulemaking hearings followed by imposition of very conservative Nuclear Regulatory Commission requirements. When evaluated on an engineering best estimate approach, the significant margins to safety for a LOCA become more apparent. Extensive test programs have also substantiated margins to safety limits. These programs have included both separate effects and systems tests. Component testing has also been performed to substantiate performance levels under adverse combinations of environmental stress. The importance of utilizing past experience and of optimizing the deployment of incremental resources is self evident. Recent safety concerns have included specific areas such

  15. PWR secondary water chemistry guidelines

    International Nuclear Information System (INIS)

    Bell, M.J.; Blomgren, J.C.; Fackelmann, J.M.

    1982-10-01

    Steam generators in pressurized water reactor (PWR) nuclear power plants have experienced tubing degradation by a variety of corrosion-related mechanisms which depend directly on secondary water chemistry. As a result of this experience, the Steam Generator Owners Group and EPRI have sponsored a major program to provide solutions to PWR steam generator problems. This report, PWR Secondary Water Chemistry Guidelines, in addition to presenting justification for water chemistry control parameters, discusses available analytical methods, data management and surveillance, and the management philosophy required to successfully implement the guidelines

  16. PWR burnable absorber evaluation

    International Nuclear Information System (INIS)

    Cacciapouti, R.J.; Weader, R.J.; Malone, J.P.

    1995-01-01

    The purpose of the study was to evaluate the relative neurotic efficiency and fuel cycle cost benefits of PWR burnable absorbers. Establishment of reference low-leakage equilibrium in-core fuel management plans for 12-, 18- and 24-month cycles. Review of the fuel management impact of the integral fuel burnable absorber (IFBA), erbium and gadolinium. Calculation of the U 3 O 8 , UF 6 , SWU, fuel fabrication, and burnable absorber requirements for the defined fuel management plans. Estimation of fuel cycle costs of each fuel management plan at spot market and long-term market fuel prices. Estimation of the comparative savings of the different burnable absorbers in dollar equivalent per kgU of fabricated fuel. (author)

  17. Steam generators in PWR's

    International Nuclear Information System (INIS)

    Michel, R.

    1974-01-01

    The steam generator of the PWR operates according to the principle of natural circulation. It consists of a U-shaped tube bundle whose free ends are welded to a bottom plate. The tube bundle is surrounded by a cylinder jacket which has slots closely above the bottom or tube plate. The feed water mixed with boiling water enters the tube bundle through these slots. Because of its buoyancy, the steam-water mixture flows upwards. Below the tube plate there are chambers for distributing and collecting pressurized water separated by means of a partition wall. By omitting some tubes, a free alloy is created so that the tubes in the center get sufficient water, too. By asymmetrical arrangement of the partition wall it is further possible to limit the tube alloy only to the inlet side for pressurized water. The flow over the tube plate is thus improved on the inlet side. (DG) [de

  18. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 9: PRAISE computer code user's manual. Final report

    International Nuclear Information System (INIS)

    Lim, E.Y.

    1981-08-01

    The PRAISE (Piping Reliability Analysis Including Seismic Events) computer code estimates the influence of earthquakes on the probability of failure at a weld joint in the primary coolant system of a pressurized water reactor. Failure, either a through-wall defect (leak) or a complete pipe severance (a large-LOCA), is assumed to be caused by fatigue crack growth of an as-fabricated interior surface circumferential defect. These defects are assumed to be two-dimensional and semi-elliptical in shape. The distribution of initial crack sizes is a function of crack depth and aspect ratio. Crack propagation rates are governed by a Paris-type relationship with separate RMS cyclic stress intensity factors for the depth and length. Both uniform through the wall and radial gradient thermal stresses are included in the calculation of the stress intensity factors. The failure probabilities are estimated by applying Monte Carlo methods to simulate the life histories of the selected weld joint. In order to maximize computational efficiency, a stratified sampling procedure is used to select the initial crack size. Hydrostatic proof test, pre-service inspection, and in-service inspection can be simulated. PRAISE treats the inter-arrival times of operating transients either as a constant or exponentially distributed according to observed or postulated rates. Leak rate and leak detection models are also included. The criterion for complete pipe severance is exceedance of a net section critical stress. Earthquakes of various intensity and arbitrary occurrence times can be modeled. PRAISE presently assumes that exactly one initial defect exists in the weld and that the earthquake of interest is the first earthquake experienced at the reactor

  19. Coolant degassing device for PWR type reactors

    International Nuclear Information System (INIS)

    Kita, Kaoru; Takezawa, Kazuaki; Minemoto, Masaki.

    1982-01-01

    Purpose: To efficiently decrease the rare gas concentration in primary coolants, as well as shorten the degassing time required for the periodical inspection in the waste gas processing system of a PWR type reactor. Constitution: Usual degassing method by supplying hydrogen or nitrogen to a volume control tank is replaced with a method of utilizing a degassing tower (method of flowing down processing liquid into the filled tower from above while uprising streams from the bottom of the tower thereby degassing the gases dissolved in the liquid into the steams). The degassing tower is combined with a hydrogen separator or hydrogen recombiner to constitute a waste gas processing system. (Ikeda, J.)

  20. Burst protected nuclear reactor plant with PWR

    International Nuclear Information System (INIS)

    Harand, E.; Michel, E.

    1978-01-01

    In the PWR, several integrated components from the steam raising unit and the main coolant pump are grouped around the reactor pressure vessel in a multiloop circuit and in a vertical arrangement. For safety reasons all primary circuit components and pipelines are situated in burst protection covers. To reduce the area of the plant straight tube steam raising units with forced circulation are used as steam raising units. The boiler pumps are connected to the vertical tubes and to the pressure vessel via double pipelines made as twin chamber pipes. (DG) [de

  1. Recent development in PWR zinc injection

    International Nuclear Information System (INIS)

    Ocken, H.; Fruzzetti, K.; Frattini, P.; Wood, C.J.

    2002-01-01

    Zinc injection to the reactor coolant system (RCS) of PWRs holds the promise to alleviate two key challenges facing PWR plant operators: (1) reducing degradation of coolant system materials, including nickel-base alloy tubing and lower alloy penetrations due to stress corrosion cracking, and (2) lowering shutdown dose rates. Primary water stress corrosion cracking (PWSCC) is a dominant tube failure mode at many plants. This paper summarizes recent observations from U. S. and international PWRs that have implemented zinc injection, focusing primarily on coolant chemistry and dose rate issues. It also provides a look at the future direction of EPRI-sponsored projects on this topic. (authors)

  2. Contribution to study and design of PWR plant simulation code

    International Nuclear Information System (INIS)

    Delourme, Didier.

    1980-11-01

    This paper presents an improvement of PICOLO, a package for PWR plants simulation. Its describes principally the integration to the code of a primary loop and pressurizer model and the corresponding control loops. Fast transients are tested on the packages and results are compared with real transients obtained on plants [fr

  3. Approximation for maximum pressure calculation in containment of PWR reactors

    International Nuclear Information System (INIS)

    Souza, A.L. de

    1989-01-01

    A correlation was developed to estimate the maximum pressure of dry containment of PWR following a Loss-of-Coolant Accident - LOCA. The expression proposed is a function of the total energy released to the containment by the primary circuit, of the free volume of the containment building and of the total surface are of the heat-conducting structures. The results show good agreement with those present in Final Safety Analysis Report - FSAR of several PWR's plants. The errors are in the order of ± 12%. (author) [pt

  4. PHEDRE model for the simulation of PWR reactors

    International Nuclear Information System (INIS)

    Bernard, Patrice; Dupraz, Remy; Vasile, Alfredo.

    1979-11-01

    This note presents the model of PHEDRE, simulator of a PWR, set on the hybrid computers of CISI, at the Nuclear Research Center of Cadarache. The model mainly concerns the primary part and the steam production of the PWR constructed in France. It includes an axial modelization of the core, the pressurizer, two loops of steam production and the inlet of the turbine, and the regulations concerning these components. The note presents the equations of the model, the structures of the codes concerning the initialization and the dynamic resolution, and describes the control panel of PHEDRE [fr

  5. 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.

  6. Transient performance of flow in circuits of PWR type reactors

    International Nuclear Information System (INIS)

    Hirdes, V.R.; Carajilescov, P.

    1988-09-01

    Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which could cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

  7. Transient performance of flow in PWR reactor circuits

    International Nuclear Information System (INIS)

    Hirdes, V.R.T.R.; Carajilescov, P.

    1988-12-01

    Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

  8. Modeling on a PWR power conversion system with system program

    International Nuclear Information System (INIS)

    Gao Rui; Yang Yanhua; Lin Meng

    2007-01-01

    Based on the power conversion system of nuclear and conventional islands of Daya Bay Power Station, this paper models the thermal-hydraulic systems of primary and secondary loops for PWR by using the PWR best-estimate program-RELAP5. To simulate the full-scope power conversion system, not only the traditional basic system models of nuclear island, but also the major system models of conventional island are all considered and modeled. A comparison between the calculated results and the actual data of reactor demonstrates a fine match for Daya Bay Nuclear Power Station, and manifests the feasibility in simulating full-scope power conversion system of PWR by RELAP5 at the same time. (authors)

  9. Measurement of the residual stresses in a PWR Control Rod Drive Mechanism nozzle

    OpenAIRE

    Coules, Harry; Smith, David

    2018-01-01

    Residual stress in the welds that attach Control Rod Drive Mechanism nozzles into the upper head of a PWR reactor vessel can influence the vessel's structural integrity and initiate Primary Water Stress Corrosion Cracking. PWSCC at Alloy 600 CRDM nozzles has caused primary coolant leakage in operating PWRs. We have used Deep Hole Drilling to characterise residual stresses in a PWR vessel head. Measurements of the internal cladding and nozzle attachment weld showed that although modest tensile...

  10. An integrated PWR for marine propulsion

    International Nuclear Information System (INIS)

    Letouze, A.; Marecaux, A.; Rollason, J.; Heap, S.; Foster, A.; Jewer, S.; Thompson, A. C.; Williams, A. M.; Beeley, P. A.

    2008-01-01

    Results from a design study for a nuclear propulsion plant utilising a small integrated PWR using many of the inherent safety features of the IRIS design. The design consists of a single pass, low enrichment core housed, together with all associated primary circuit components, within a reactor pressure vessel 10.3 m high and 4.1 m in diameter. Reactor physics calculations were conducted with the codes WIMS9a and MONK8b. The core design contains 21 fuel assemblies each containing 264 UO 2 fuel pins. Each fuel module has a cluster of 24 boron carbide control rods and a central instrumentation channel. The fuel enrichment was 9% in order to achieve the core lifetime requirement of 3000 EFPD at a reactor power of 120 MWth. This gives a discharge burnup of 51,000 MWd/t. To control excess reactivity, two forms of burnable poison are employed: a zirconium dibromide (ZrB 2 ) coating on the fuel compacts, and gadolinium oxide homogeneously mixed in the fuel. Thermal hydraulic calculations were performed using TRAC-P(ND) for steady-state operation and for a number of fault transients. The helical once through steam generators were modelled using heat structure and pipe components and their performance compared to independent calculations including heat transfer correlations for the helical coiled geometry. Intact circuit calculations for steady state were followed by a small break LOCA calculation including the effect of a containment volume which reproduced the gain of coolant effect reported for IRIS. It was demonstrated that the thermal limits were not exceeded for the identified key transients. The dynamic response of the reactor plant to typical power demands was modelled using AcslXtreme software. Several schemes for limiting the power overshoot that was found on rapid increase to full power were examined. It was concluded that the SG must be operated with variable secondary pressure and the best means of reducing power overshoot is to step back the throttle opening

  11. On site PWR fuel inspection measurements for operational and design verification

    International Nuclear Information System (INIS)

    1996-01-01

    The on-site inspection of irradiated Pressurized Water Reactor (PWR) fuel and Non-Fuel Bearing Components (NFBC) is typically limited to visual inspections during refuelings using underwater TV cameras and is intended primarily to confirm whether the components will continue in operation. These inspections do not normally provide data for design verification nor information to benefit future fuel designs. Japanese PWR utilities and Nuclear Fuel Industries Ltd. designed, built, and performed demonstration tests of on-site inspection equipment that confirms operational readiness of PWR fuel and NFBC and also gathers data for design verification of these components. 4 figs, 3 tabs

  12. French PWR Safety Philosophy

    International Nuclear Information System (INIS)

    Conte, M. M.

    1986-01-01

    The first 900 MWe units, built under the American Westinghouse licence and with reference to the U. S. regulation, were followed by 28 standardized units, C P1 and C P2 series. Increasing knowledge and lessons learned from starting and operating experience of French nuclear power plants, completed by the experience learned from the operation of foreign reactors, has contributed to the improvement of French PWR design and safety philosophy. As early as 1976, this experience was taken into account by French Safety organisms to discuss, with Electricite de France, the safety options for the planned 1300 MWe units, P4 and P4 series. In 1983, the new reactor scheduled, Ni4 series 1400 MWe, is a totally French design which satisfies the French regulations and other French standards and codes. Based on a deterministic approach, the French safety analysis was progressively completed by a probabilistic approach each of them having possibilities and limits. Increasing knowledge and lessons learned from operating experience have contributed to the French safety philosophy improvement. The methodology now applied to safety evaluation develops a new facet of the in depth defense concept by taking highly unlikely events into consideration, by developing the search of safety consistency of the design, and by completing the deterministic approach by the probabilistic one

  13. Resfria - a computational routine for thermal-hydraulic analysis of a cooldown in the PWR

    International Nuclear Information System (INIS)

    Silva Neto, A.J. da; Maciel Filho, L.A.

    1989-01-01

    This paper presents the computer code RESFRIA, designed to calculate the process parameters in a PWR nuclear power plant during a cooldown normal procedure. The procedure is described and some of the models developed to the simulation of systems and equipments are presented. A simplified flowchart of the computational routine and the results in the form of a diagram, for a typical PWR nuclear power plant, are also presented. (author)

  14. Modeling of PWR fuel at extended burnup

    International Nuclear Information System (INIS)

    Dias, Raphael Mejias

    2016-01-01

    This work studies the modifications implemented over successive versions in the empirical models of the computer program FRAPCON used to simulate the steady state irradiation performance of Pressurized Water Reactor (PWR) fuel rods under high burnup condition. In the study, the empirical models present in FRAPCON official documentation were analyzed. A literature study was conducted on the effects of high burnup in nuclear fuels and to improve the understanding of the models used by FRAPCON program in these conditions. A steady state fuel performance analysis was conducted for a typical PWR fuel rod using FRAPCON program versions 3.3, 3.4, and 3.5. The results presented by the different versions of the program were compared in order to verify the impact of model changes in the output parameters of the program. It was observed that the changes brought significant differences in the results of the fuel rod thermal and mechanical parameters, especially when they evolved from FRAPCON-3.3 version to FRAPCON-3.5 version. Lower temperatures, lower cladding stress and strain, lower cladding oxide layer thickness were obtained in the fuel rod analyzed with the FRAPCON-3.5 version. (author)

  15. Valve testing for UK PWR safety applications

    International Nuclear Information System (INIS)

    George, P.T.; Bryant, S.

    1989-01-01

    Extensive testing and development has been done by the Central Electricity Generating Board (CEGB) to support the design, construction and operation of Sizewell B, the UK's first PWR. A Blowdown Rig for the Assessment of Valve Operability - (BRAVO) has been constructed at the CEGB Marchwood Engineering Laboratory to reproduce PWR Pressurizer fluid conditions for the full scale testing of Pressurizer Relief System (PRS) valves. A full size tandem pair of Pilot Operated Safety Relief Valves (POSRVs) is being tested under the full range of pressurizer fluid conditions. Tests to date have produced important data on the performance of the valve in its Cold Overpressure protection mode of operation and on methods for the in-service testing of the valve. Also, a full size pressurizer safety valve has been tested under full PRS fluid conditions to develop a methodology for the pre-service testing of the Sizewell valves. Further work will be carried out to develop procedures for the in-service testing of the valve. In the Main Steam Safety Valve test program carried out at the Siemens-KWU Test Facilities, a single MSSV from three potential suppliers was tested under full secondary system conditions. The test results have been analyzed and are reflected in the CEGB's arrangements for the pre-service and in-service testing of the Sizewell MSSVs. Valves required to interrupt pipebreak flow must be qualified for this duty by testing or a combination of testing and analysis. To obtain guidance on the performance of such tests gate and globe valves have been subjected to simulated pipebreaks under PWR primary circuit conditions. In the light of problems encountered with gate valve closure under these conditions, further tests are currently being carried out on the BRAVO facility on a gate valve, in preparation for the full scale flow interruption qualification testing of the Sizewell main steam isolation valve

  16. 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)

  17. Implementation in free software of the PWR type university nucleo electric simulator (SU-PWR); Implementacion en software libre del simulador universitario de nucleoelectrica tipo PWR (SU-PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Hidago H, F.; Morales S, J.B. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: julfi_jg@yahoo.com.mx

    2007-07-01

    Presently work is shown like was carried out the implementation of the University Simulator of Nucleo-electric type PWR (SU-PWR). The implementation of the simulator was carried out in a free software simulation platform, as it is Scilab, what offers big advantages that go from the free use and without cost of the product, until the codes modification so much of the system like of the program with the purpose of to improve it or to adapt it to future routines and/or more advanced graphic interfaces. The SU-PWR shows the general behavior of a PWR nuclear plant (Pressurized Water Reactor) describing the dynamics of the plant from the generation process of thermal energy in the nuclear fuel, going by the process of energy transport toward the coolant of the primary circuit the one which in turn transfers this energy to the vapor generators of the secondary circuit where the vapor is expanded by means of turbines that in turn move the electric generator producing in this way the electricity. The pressurizer that is indispensable for the process is also modeled. Each one of these stages were implemented in scicos that is the Scilab tool specialized in the simulation. The simulation was carried out by means of modules that contain the differential equation that mathematically models each stage or equipment of the PWR plant. The result is a series of modules that based on certain entrances and characteristic of the system they generate exits that in turn are the entrance to other module. Because the SU-PWR is an experimental project in early phase, it is even work and modifications to carry out, for what the models that are presented in this work can vary a little the being integrated to the whole system to simulate, but however they already show clearly the operation and the conformation of the plant. (Author)

  18. Sizewell 'B' PWR reference design

    International Nuclear Information System (INIS)

    1982-04-01

    The reference design for a PWR power station to be constructed as Sizewell 'B' is presented in 3 volumes containing 14 chapters and in a volume of drawings. The report describes the proposed design and provides the basis upon which the safety case and the Pre-Construction Safety Report have been prepared. The station is based on a 3425MWt Westinghouse PWR providing steam to two turbine generators each of 600 MW. The layout and many of the systems are based on the SNUPPS design for Callaway which has been chosen as the US reference plant for the project. (U.K.)

  19. 14C Behaviour in PWR coolant

    International Nuclear Information System (INIS)

    Sims, Howard; Dickinson Shirley; Garbett, Keith

    2012-09-01

    Although 14 C is produced in relatively small amounts in PWR coolant, it is important to know its fate, for example whether it is released by gaseous discharge, removed by absorption on ion exchange (IX) resins or deposited on the fuel pin surfaces. 14 C can exist in a range of possible chemical forms: inorganic carbon compounds (probably mainly CO 2 ), elemental carbon, and organic compounds such as hydrocarbons. This paper presents results from a preliminary survey of the possible reactions of 14 C in PWR coolant. The main conclusions of the study are: - A combination of thermal and radiolytic reactions controls the chemistry of 14 C in reactor coolant. A simple chemical kinetic model predicts that CH 3 OH would be the initial product from radiolytic reactions of 14 C following its formation from 17 O. CH 3 OH is predicted to arise as a result of reactions of OH . with CH 4 and CH 3 , and it persists because there is no known radiation chemical reduction mechanism. - Thermodynamic considerations show that CH 3 OH can be thermally reduced to CH 4 in PWR conditions, although formation of CO 2 from small organics is the most thermodynamically favourable outcome. Such reactions could be catalysed on active nickel surfaces in the primary circuit. - Limited plant data would suggest that CH 4 is the dominant form in PWR and CO 2 in BWR. This implies that radiation chemistry may be important in determining the speciation. - Addition of acetate does not affect the amount of 14 C formed, but the addition of large amounts of stable carbon would lead to a large range of additional products, some of which would be expected to deposit on fuel pin surfaces as high molecular weight hydrocarbons. However, the subsequent thermal decomposition reactions of these products are not known. - Acetate addition may represent a small input of 12 C compared with organic material released from CVCS resins, although the importance of this may depend on whether that is predominantly soluble

  20. Effects of Burnable Absorbers on PWR Spent Nuclear Fuel

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  1. Advanced high conversion PWR: preliminary analysis

    International Nuclear Information System (INIS)

    Golfier, H.; Bellanger, V.; Bergeron, A.; Dolci, F.; Gastaldi, B.; Koberl, O.; Mignot, G.; Thevenot, C.

    2007-01-01

    In this paper, physical aspects of a HCPWR (High Conversion Light Water Reactor), which is an innovative PWR fuelled with mixed oxide and having a higher conversion ratio due to a lower moderation ratio. Moderation ratios lower than unity are considered which has led to low moderation PWR fuel assembly designs. The objectives of this parametric study are to define a feasibility area with regard to the following neutronic aspects: moderation ratio, Pu loading, reactor spectrum, irradiation time, and neutronic coefficients. Important thermohydraulic parameters are the pressure drop, the critical heat flux, the maximum temperature in the fuel rod and the pumping power. The thermohydraulic analysis shows that a range of moderation ratios from 0.8 to 1.2 is technically possible. A compromise between improved fuel utilization and research and development effort has been found for the moderation ration of about 1. The parametric study shows that there are 2 ranges of interest for the moderation ratio: -) moderation ratio between 0.8 and 1.2 with reduced fissile heights (> 3 m), hexagonal arrangement fuel assembly and square arrangement fuel assembly are possible; and -) moderation between 0.6 and 0.7 with a modification of the reactor operating conditions (reduction of the primary flow and of the thermal power), the fuel rods could be arranged inside a hexagonal fuel rod assembly. (A.C.)

  2. Influences of boric acid and lithium hydroxide on oxide film of type 316 stainless steel in PWR simulated primary water; PWR 1次冷却材模擬環境中の316ステンレス鋼に生成した皮膜性状に及ぼすほう酸および水酸化リチウムの影響

    Energy Technology Data Exchange (ETDEWEB)

    Fukumura, Takuya; Fukuya, Koji; Arioka, Koji [Institute of Nuclear Safety System, Inc., Mihama, Fukui (Japan)

    2012-06-15

    In order to understand the influences of boric acid and lithium hydroxide on the IGSCC of type 316 stainless steel, an oxide film was analyzed in simulated PWR primary water while varying the boric acid and lithium hydroxide concentrations. It was found that, although boric acid and lithium hydroxide did not affect the structure and chemical composition of the surface oxide film remarkably, a lower boric acid concentration or a higher lithium concentration produced an oxide film with a thicker surface. It was considered that the lower boric acid concentration and higher lithium hydroxide concentration caused a higher magnetite solubility at the surface of the material and that the higher magnetite solubility caused a higher iron concentration gradient, which promoted iron dissolution from the material and the formation of a thicker oxide film. It was found that the thicker oxide film caused a higher IGSCC susceptibility and that the corrosion was the dominant factor of the IGSCC mechanism. No significant change was found in the morphologies of crack tip oxide in different bulk water chemistry systems, thus producing CT specimens with similar crack growth rates. (author)

  3. Study on mitigation of in-vessel release of fission products in severe accidents of PWR

    International Nuclear Information System (INIS)

    Huang, G.F.; Tong, L.L.; Li, J.X.; Cao, X.W.

    2010-01-01

    Research highlights: → In-vessel release of fission products in severe accidents for 600 MW PWR is analyzed. → Mitigation effect of primary feed-and-bleed on in-vessel release is investigated. → Mitigation effect of secondary feed-and-bleed on in-vessel release is studied. → Mitigation effect of ex-vessel cooling on in-vessel release is evaluated. - Abstract: During the severe accidents in a nuclear power plant, large amounts of fission products release with accident progression, including in-vessel and ex-vessel release. Mitigation of fission products release is demanded for alleviating radiological consequence in severe accidents. Mitigation countermeasures to in-vessel release are studied for Chinese 600 MW pressurized water reactor (PWR), including feed-and-bleed in primary circuit, feed-and-bleed in secondary circuit and ex-vessel cooling. SBO, LOFW, SBLOCA and LBLOCA are selected as typical severe accident sequences. Based on the evaluation of in-vessel release with different startup time of countermeasure, and the coupling relationship between thermohydraulics and in-vessel release of fission products, some results are achieved. Feed-and-bleed in primary circuit is an effective countermeasure to mitigate in-vessel release of fission products, and earlier startup time of countermeasure is more feasible. Feed-and-bleed in secondary circuit is also an effective countermeasure to mitigate in-vessel release for most severe accident sequences that can cease core melt progression, e.g. SBO, LOFW and SBLOCA. Ex-vessel cooling has no mitigation effect on in-vessel release owing to inevitable core melt and relocation.

  4. Effect of water chemistry on deposition for PWR plant operation

    International Nuclear Information System (INIS)

    Le Calvar, Marc; Bretelle, J. L.; Cailleaux, J. P.; Lacroix, R.; Guivarch, M.; Gay, N.; Taunier, S.; Gressier, F.; Varry, P.; Corredera, G.; Alos-Ramos, O.; Dijoux, M.

    2012-09-01

    For Pressurized Water Reactor (PWR) operation, water chemistry guidelines, specifications and associated surveillance programs are key to avoid deposition of oxides. Deposition of oxides can be detrimental by disrupting results of flow measurements, decreasing the thermal exchange capacity, or even by impairing safety. This paper describes the most important cases of deposition, their consequences for operation, and the implemented improvements to avoid their reoccurrence. Deposition that led to a Crud Induced Power Shift (CIPS) is also described. In the primary and in the secondary sides, orifice plates are typically used for measuring feedwater flow rate in nuclear power plants. Feedwater flow rates are used for control purposes and are important safety parameters as they are used to determine the plant's operating power level. Fouling of orifice plates in the primary side has been found during surveillance testing. For reactor coolant pumps, the formation of deposits on the seal No.1 can cause abnormally high or low leak rates through the seal. The leak rate through this seal must be carefully maintained within a prescribed range during plant operation. In the secondary side, orifice plate fouling has been the cause of feedwater flow/reference thermal power drift. For the steam generators (SG), magnetite deposition has led to fouling of the tube bundle, clogging of the quadri-foiled support plate holes and hard sludge formation on the base plate. For the generators, copper hollow conductors are widely used. Buildup of copper oxides on the interior walls of copper conductors has caused insufficient heat transfer. All these deposition cases have received adequate attention, understanding and response via improvement of our surveillance programs. (authors)

  5. PWR-to-PWR fuel cycle model using dry process

    International Nuclear Information System (INIS)

    Iqbal, M.; Jeong, Chang Joon; Rho, Gyu Hong

    2002-03-01

    PWR-to-PWR fuel cycle model has been developed to recycle the spent fuel using the dry fabrication process. Two types of fuels were considered; first fuel was based on low initial enrichment with low discharge burnup and second one was based on more initial enrichment with high discharge burnup in PWR. For recycling calculations, the HELIOS code was used, in which all of the available fission products were considered. The decay of 10 years was applied for reuse of the spent fuel. Sensitivity analysis for the fresh feed material enrichment has also been carried out. If enrichment of the mixing material is increased the saving of uranium reserves would be decreased. The uranium saving of low burned fuel increased from 4.2% to 7.4% in fifth recycling step for 5 wt% to 19.00wt% mixing material enrichment. While for high burned fuel, there was no uranium saving, which implies that higher uranium enrichment required than 5 wt%. For mixing of 15 wt% enriched fuel, the required mixing is about 21.0% and 37.0% of total fuel volume for low and high burned fuel, respectively. With multiple recycling, reductions in waste for low and high burned fuel became 80% and 60%, for first recycling, respectively. In this way, waste can be reduced more and the cost of the waste disposal reduction can provide the economic balance

  6. Gadolinia experience and design for PWR fuel cycles

    International Nuclear Information System (INIS)

    Stephenson, L. C.

    2000-01-01

    The purpose of this paper is to describe Siemens Power Corporation's (SPC) current experience with the burnable absorber gadolinia in PWR fuel assemblies, including optimized features of SPC's PWR gadolinia designs, and comparisons with other burnable absorbers. Siemens is the world leader in PWR gadolinia experience. More than 5,900 Siemens PWR gadolinia-bearing fuel assemblies have been irradiated. The use of gadolinia-bearing fuel provides significant flexibility in fuel cycle designs, allows for low radial leakage fuel management and extended operating cycles, and reduces BOC (beginning-of-cycle) soluble boron concentrations. The optimized use of an integral burnable neutron absorber is a design feature which provides improved economic performance for PWR fuel assemblies. This paper includes a comparison between three different types of integral burnable absorbers: gadolinia, Zirconium diboride and erbia. Fuel cycle design studies performed by Siemens have shown that the enrichment requirements for 18-24 month fuel cycles utilizing gadolinia or zirconium diboride integral fuel burnable absorbers can be approximately the same. Although a typical gadolinia residual penalty for a cycle design of this length is as low as 0.02-0.03 wt% U-235, the design flexibility of gadolinia allows for very aggressive low-leakage core loading plans which reduces the enrichment requirements for gadolinia-bearing fuel. SPC has optimized its use of gadolinia in PWR fuel cycles. Typically, low (2-4) weight percent Gd 2 O 3 is used for beginning to middle of cycle reactivity hold down as well as soluble boron concentration holddown at BOC. Higher concentrations of Gd 2 O 3 , such as 6 and 8 wt%, are used to control power peaking in assemblies later in the cycle. SPC has developed core strategies that maximize the use of lower gadolinia concentrations which significantly reduces the gadolinia residual reactivity penalty. This optimization includes minimizing the number of rods with

  7. A nodal model for the simulation of a PWR core

    International Nuclear Information System (INIS)

    Souza Pinto, R. de.

    1981-06-01

    A computer program FORTRAN language was developed to simulate the neutronic and thermal-hydraulic transient behaviour of a PWR reactor core. The reator power is calculated using a point kinectics model with six groups of delayed neutron precursors. The fission product decay heat was considered assuming three effective decay heat groups. A nodal model was employed for the treatment of heat transfer in the fuel rod, with integration of the heat equation by the lumped parameter technique. Axial conduction was neglected. A single-channel nodal model was developed for the thermo-hydrodynamic simulation using mass and energy conservation equations for the control volumes. The effect of the axial pressure variation was neglected. The computer program was tested, with good results, through the simulation of the transient behaviour of postulated accidents in a typical PWR. (Author) [pt

  8. The increase in fatigue crack growth rates observed for Zircaloy-4 in a PWR environment

    Science.gov (United States)

    Cockeram, B. V.; Kammenzind, B. F.

    2018-02-01

    Cyclic stresses produced during the operation of nuclear reactors can result in the extension of cracks by processes of fatigue. Although fatigue crack growth rate (FCGR) data for Zircaloy-4 in air are available, little testing has been performed in a PWR primary water environment. Test programs have been performed by Gee et al., in 1989 and Picker and Pickles in 1984 by the UK Atomic Energy Authority, and by Wisner et al., in 1994, that have shown an enhancement in FCGR for Zircaloy-2 and Zircaloy-4 in high-temperature water. In this work, FCGR testing is performed on Zircaloy-4 in a PWR environment in the hydrided and non-hydrided condition over a range of stress-intensity. Measurements of crack extension are performed using a direct current potential drop (DCPD) method. The cyclic rate in the PWR primary water environment is varied between 1 cycle per minute to 0.1 cycle per minute. Faster FCGR rates are observed in water in comparison to FCGR testing performed in air for the hydrided material. Hydrided and non-hydrided materials had similar FCGR values in air, but the non-hydrided material exhibited much lower rates of FCGR in a PWR primary water environment than for hydrided material. Hydrides are shown to exhibit an increased tendency for cracking or decohesion in a PWR primary water environment that results in an enhancement in FCGR values. The FCGR in the PWR primary water only increased slightly with decreasing cycle frequency in the range of 1 cycle per minute to 0.1 cycle per minute. Comparisons between the FCGR in water and air show the enhancement from the PWR environment is affected by the applied stress intensity.

  9. PWR plant construction in Japan

    International Nuclear Information System (INIS)

    Tamura, Toshifumi

    2002-01-01

    The construction methods based on the experiences on the Nuclear Island, which is a critical path in the total construction schedule, have been studied and reconsidered in order to construct by more reliable and economical method. So various improved construction method are being applied and the duration of construction is being reduced continuously. So various improved construction method are being applied and the duration of construction is being reduced continuously. In this paper, the history of construction of twenty-three (23) PWR Plant, the actual construction methods and schedule of Ohi-3/4, to which the many improved methods were applied during their construction, are introduced mainly with the improved points for previously constructed plants. And also the situation of construction method for the next PWR Plant is simply explained

  10. A comparison of fuzzy logic-PID control strategies for PWR pressurizer control

    International Nuclear Information System (INIS)

    Kavaklioglu, K.; Ikonomopoulos, A.

    1993-01-01

    This paper describes the results obtained from a comparison performed between classical proportional-integral-derivative (PID) and fuzzy logic (FL) controlling the pressure in a pressurized water reactor (PWR). The two methodologies have been tested under various transient scenarios, and their performances are evaluated with respect to robustness and on-time response to external stimuli. One of the main concerns in the safe operation of PWR is the pressure control in the primary side of the system. In order to maintain the pressure in a PWR at the desired level, the pressurizer component equipped with sprayers, heaters, and safety relief valves is used. The control strategy in a Westinghouse PWR is implemented with a PID controller that initiates either the electric heaters or the sprayers, depending on the direction of the coolant pressure deviation from the setpoint

  11. Corrosion of PWR steam generators

    International Nuclear Information System (INIS)

    Garnsey, R.

    1979-01-01

    Some designs of pressurized water reactor (PWR) steam generators have experienced a variety of corrosion problems which include stress corrosion cracking, tube thinning, pitting, fatigue, erosion-corrosion and support plate corrosion resulting in 'denting'. Large international research programmes have been mounted to investigate the phenomena. The operational experience is reviewed and mechanisms which have been proposed to explain the corrosion damage are presented. The implications for design development and for boiler and feedwater control are discussed. (author)

  12. PWR system reliability improvement activities

    International Nuclear Information System (INIS)

    Yoshikawa, Yuichiro

    1985-01-01

    In Japan lacking in energy resources, it is our basic energy policy to accelerate the development program of nuclear power, thereby reducing our dependence. As referred to in the foregoing, every effort has been exerted on our part to improve the PWR system reliability by dint of the so-called 'HOMEMADE' TQC activities, which is our brain-child as a result of applying to the energy industry the quality control philosophy developed in the field of manufacturing industry

  13. Sun-induced chlorophyll fluorescence and photochemical reflectance index improve remote-sensing gross primary production estimates under varying nutrient availability in a typical Mediterranean savanna ecosystem

    Science.gov (United States)

    Perez-Priego, O.; Guan, J.; Rossini, M.; Fava, F.; Wutzler, T.; Moreno, G.; Carvalhais, N.; Carrara, A.; Kolle, O.; Julitta, T.; Schrumpf, M.; Reichstein, M.; Migliavacca, M.

    2015-11-01

    This study investigates the performances of different optical indices to estimate gross primary production (GPP) of herbaceous stratum in a Mediterranean savanna with different nitrogen (N) and phosphorous (P) availability. Sun-induced chlorophyll fluorescence yield computed at 760 nm (Fy760), scaled photochemical reflectance index (sPRI), MERIS terrestrial-chlorophyll index (MTCI) and normalized difference vegetation index (NDVI) were computed from near-surface field spectroscopy measurements collected using high spectral resolution spectrometers covering the visible near-infrared regions. GPP was measured using canopy chambers on the same locations sampled by the spectrometers. We tested whether light-use efficiency (LUE) models driven by remote-sensing quantities (RSMs) can better track changes in GPP caused by nutrient supplies compared to those driven exclusively by meteorological data (MM). Particularly, we compared the performances of different RSM formulations - relying on the use of Fy760 or sPRI as a proxy for LUE and NDVI or MTCI as a fraction of absorbed photosynthetically active radiation (fAPAR) - with those of classical MM. Results showed higher GPP in the N-fertilized experimental plots during the growing period. These differences in GPP disappeared in the drying period when senescence effects masked out potential differences due to plant N content. Consequently, although MTCI was closely related to the mean of plant N content across treatments (r2 = 0.86, p < 0.01), it was poorly related to GPP (r2 = 0.45, p < 0.05). On the contrary sPRI and Fy760 correlated well with GPP during the whole measurement period. Results revealed that the relationship between GPP and Fy760 is not unique across treatments, but it is affected by N availability. Results from a cross-validation analysis showed that MM (AICcv = 127, MEcv = 0.879) outperformed RSM (AICcv =140, MEcv = 0.8737) when soil moisture was used to constrain the seasonal dynamic of LUE. However

  14. PWR secondary water chemistry guidelines: Revision 3

    International Nuclear Information System (INIS)

    Lurie, S.; Bucci, G.; Johnson, L.; King, M.; Lamanna, L.; Morgan, E.; Bates, J.; Burns, R.; Eaker, R.; Ward, G.; Linnenbom, V.; Millet, P.; Paine, J.P.; Wood, C.J.; Gatten, T.; Meatheany, D.; Seager, J.; Thompson, R.; Brobst, G.; Connor, W.; Lewis, G.; Shirmer, R.; Gillen, J.; Kerns, M.; Jones, V.; Lappegaard, S.; Sawochka, S.; Smith, F.; Spires, D.; Pagan, S.; Gardner, J.; Polidoroff, T.; Lambert, S.; Dahl, B.; Hundley, F.; Miller, B.; Andersson, P.; Briden, D.; Fellers, B.; Harvey, S.; Polchow, J.; Rootham, M.; Fredrichs, T.; Flint, W.

    1993-05-01

    An effective, state-of-the art secondary water chemistry control program is essential to maximize the availability and operating life of major PWR components. Furthermore, the costs related to maintaining secondary water chemistry will likely be less than the repair or replacement of steam generators or large turbine rotors, with resulting outages taken into account. The revised PWR secondary water chemistry guidelines in this report represent the latest field and laboratory data on steam generator corrosion phenomena. This document supersedes Interim PWR Secondary Water Chemistry Recommendations for IGA/SCC Control (EPRI report TR-101230) as well as PWR Secondary Water Chemistry Guidelines--Revision 2 (NP-6239)

  15. Economic optimization of PWR cores with ROSA

    International Nuclear Information System (INIS)

    Verhagen, F.C.M.; Wakker, P.H.

    2005-01-01

    The core-loading pattern is decisive for fuel cycle economics, fuel safety parameters and economic planning for future cycles. ROSA, NRG's loading pattern optimization code system for PWRs, has proven for over a decade to be a valuable tool to reactor operators for improving their fuel management economics. ROSA uses simulated annealing as loading pattern optimization technique, in combination with an extremely fast 3-D neutronics code for loading pattern calculations. The code is continuously extended with new optimization parameters and rules. This paper outlines recent developments of the ROSA code system and discusses results of PWR specific applications of ROSA. Core designs with a large variety of challenging constraints have been realized with ROSA. As a typical example, for the 193 assembly, Vantage 5H/RFA-2 fueled TVA's Watts Bar unit 1, a cycle 4 core with 76 feed assemblies was designed. This was followed by a high-energy cycle 5 with only 77 feed assemblies and approximately 535 days of natural cycle length. Subsequently, an economical core using 72 bundles was designed for cycle 6. This resulted in considerable savings in the cost of feed assemblies for reloads. The typical accuracy of ROSA compared to results of license codes in within ±0.02 for normalized assembly powers, ±0.03 for maximum enthalpy rise hot channel factor (F ΔH ), and ±3 days for natural cycle length. (author)

  16. PWR plant transient analyses using TRAC-PF1

    International Nuclear Information System (INIS)

    Ireland, J.R.; Boyack, B.E.

    1984-01-01

    This paper describes some of the pressurized water reactor (PWR) transient analyses performed at Los Alamos for the US Nuclear Regulatory Commission using the Transient Reactor Analysis Code (TRAC-PF1). Many of the transient analyses performed directly address current PWR safety issues. Included in this paper are examples of two safety issues addressed by TRAC-PF1. These examples are pressurized thermal shock (PTS) and feed-and-bleed cooling for Oconee-1. The calculations performed were plant specific in that details of both the primary and secondary sides were modeled in addition to models of the plant integrated control systems. The results of these analyses show that for these two transients, the reactor cores remained covered and cooled at all times posing no real threat to the reactor system nor to the public

  17. Material property changes of stainless steels under PWR irradiation

    International Nuclear Information System (INIS)

    Fukuya, Koji; Nishioka, Hiromasa; Fujii, Katsuhiko; Kamaya, Masayuki; Miura, Terumitsu; Torimaru, Tadahiko

    2009-01-01

    Structural integrity of core structural materials is one of the key issues for long and safe operation of pressurized water reactors. The stainless steel components are exposed to neutron irradiation and high-temperature water, which cause significant property changes and irradiation assisted stress corrosion cracking (IASCC) in some cases. Understanding of irradiation induced material property changes is essential to predict integrity of core components. In the present study, microstructure and microchemistry, mechanical properties, and IASCC behavior were examined in 316 stainless steels irradiated to 1 - 73 dpa in a PWR. Dose-dependent changes of dislocation loops and cavities, grain boundary segregation, tensile properties and fracture mode, deformation behavior, and their interrelation were discussed. Tensile properties and deformation behavior were well coincident with microstructural changes. IASCC susceptibility under slow strain rate tensile tests, IASCC initiation under constant load tests in simulated PWR primary water, and their relationship to material changes were discussed. (author)

  18. Industrial assessment of nonbackfittable PWR design modifications. Final report

    International Nuclear Information System (INIS)

    Matzie, R.A.; Daleas, R.S.; Miller, D.D.

    1980-11-01

    As part of the US Department of Energy's Advanced Reactor Design Study, various nonbackfittable PWR design modifications were evaluated to determine their potential for improved uranium utilization and commercial viability. Combustion Engineering, Inc. contributed to this effort through participation in the Battelle Pacific Northwest Laboratory industrial assessment of such design modifications. Seven modifications, including the use of higher primary system temperatures and pressures, rapid-frequent refueling, end-of-cycle stretchout, core periphery modifications, radial blankets, low power density cores, and small PWR assemblies, were evaluated with respect to uranium utilization, economics, technical and operational complexity, and several other subjective considerations. Rapid-frequent refueling was judged to have the highest potential although it would probably not be economical for the majority of reactors with the design assumptions used in this assessment

  19. PWR type reactor plant

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi.

    1993-01-01

    A water chamber of a horizontal U-shaped pipe type steam generator is partitioned to an upper high temperature water chamber portion and a lower low temperature water chamber portion. An exit nozzle of a reactor container containing a reactor core therein is connected to a suction port of a coolant pump by way of first high temperature pipelines. The exit port of the coolant pump is connected to the high temperature water chamber portion of the steam generator by way of second high temperature pipelines. The low temperature water chamber portion of the steam generator is connected to an inlet nozzle of the reactor container by way of the low temperature pipelines. The low temperature water chamber portion of the steam generator is positioned lower than the high temperature water chamber portion, but upper than the reactor core. Accordingly, all of the steam generator for a primary coolant system, coolant pumps as well as high temperature pipelines and low temperature pipelines connecting them are disposed above the reactor core. With such a constitution, there is no worry of interrupting core cooling even upon occurrence of an accident, to improve plant safety. (I.N.)

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

    International Nuclear Information System (INIS)

    Claro, Luiz H.

    2009-01-01

    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 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∞ 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)

  1. Conceptual design of simplified PWR

    International Nuclear Information System (INIS)

    Tabata, Hiroaki

    1996-01-01

    The limited availability for location of nuclear power plant in Japan makes plants with higher power ratings more desirable. Having no intention of constructing medium-sized plants as a next generation standard plant, Japanese utilities are interested in applying passive technologies to large ones. So, Japanese utilities have studied large passive plants based on AP600 and SBWR as alternative future LWRs. In a joint effort to develop a new generation nuclear power plant which is more friendly to operator and maintenance personnel and is economically competitive with alternative sources of power generation, JAPC and Japanese Utilities started the study to modify AP600 and SBWR, in order to accommodate the Japanese requirements. During a six year program up to 1994, basic concepts for 1000 MWe class Simplified PWR (SPWR) and Simplified BWR (SBWR) were developed, though there still remain several areas to be improved. These studies have now stepped into the phase of reducing construction cost and searching for maximum power rating that can be attained by reasonably practical technology. These results also suggest that it is hopeful to develop a large 3-loop passive plant (∼1200 MWe). Since Korea mainly deals with PWR, this paper summarizes SPWR study. The SPWR is jointly studied by JAPC, Japanese PWR Utilities, EdF, WH and Mitsubishi Heavy Industry. Using the AP-600 reference design as a basis, we enlarged the plant size to 3-loops and added engineering features to conform with Japanese practice and Utilities' preference. The SPWR program definitively confirmed the feasibility of a passive plant with an NSSS rating about 1000 MWe and 3 loops. (J.P.N.)

  2. Conversion ratio and consumption of fissile material in PWR reactors

    International Nuclear Information System (INIS)

    Tiba, C.

    1977-01-01

    It has been shown that the uranium resources will be insufficient for future projected demand. The many solutions to this problem are considered and, in particular, the effect of enrichment on the conversion ratio and hence total uranium comsumption is studied. The developed computacional method employs the one-group neutron diffusion theory. The model is verified by calculating typical burn-up, conversion ratio, U-235 comsumption and plutonium production values in PWR's, and comparing results with those in the published literature. The associated costs of U and U-Pu fuel cycles are also studied for various enrichment values [pt

  3. Surveillance of vibrations in PWR

    International Nuclear Information System (INIS)

    Espefaelt, R.; Lorenzen, J.; Aakerhielm, F.

    1980-07-01

    The core of a PWR - including fuel elements, internal structure, control rods and core support structure inside the pressure vessel - is subjected to forces which can cause vibrations. One sensitive means to detect and analyse such vibrations is by means of the noise from incore and excore neutron detector signals. In this project noise recordings have been made on two occasions in the Ringhals 2 plant and the obtained data been analysed using the Studsvik Noise Analysis Program System (SNAPS). The results have been intepreted and a detailed description of the vibrational status of the core and pressure vessel internals has been produced. On the basis of the obtained results it is proposed that neutron signal noise analysis should be performed at each PWR plant in the beginning, middle and end of each fuel cycle and an analysis be made using the methods developed in the project. It would also provide a contribution to a higher degree of preparedness for diagnostic tasks in case of unexpected and abnormal events. (author)

  4. Acceptance test for 900 MWe PWR unit replacement steam generators

    International Nuclear Information System (INIS)

    Gourguechon, B.

    1993-01-01

    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

  5. 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.

  6. Sensitivity Verification of PWR Monitoring System Using Neuro-Expert For LOCA Detection

    International Nuclear Information System (INIS)

    Muhammad Subekti

    2009-01-01

    Sensitivity Verification of PWR Monitoring System Using Neuro-Expert For LOCA Detection. The present research was done for verification of previous developed method on Loss of Coolant Accident (LOCA) detection and perform simulations for knowing the sensitivity of the PWR monitoring system that applied neuro-expert method. The previous research continuing on present research, has developed and has tested the neuro-expert method for several anomaly detections in Nuclear Power Plant (NPP) typed Pressurized Water Reactor (PWR). Neuro-expert can detect the LOCA anomaly with sensitivity of primary coolant leakage of 7 gallon/min and the conventional method could not detect the primary coolant leakage of 30 gallon/min. Neuro expert method detects significantly LOCA anomaly faster than conventional system in Surry-1 NPP as well so that the impact risk is reducible. (author)

  7. Neutronic feasibility of PWR core with mixed oxide fuels in the Republic of Korea

    International Nuclear Information System (INIS)

    Kim, Y.J.; Joo, H.K.; Jung, H.G.; Sohn, D.S.

    1997-01-01

    Neutronic feasibility of a PWR core with mixed oxide (MOX) fuels has been investigated as part of the feasibility study for recycling spent fuels in Korea. A typical 3-loop PWR with 900 MWe capacity is selected as reference plant to develop equilibrium core designs with low-leakage fuel management scheme, while incorporating various MOX loading. The fuel management analyses and limited safety analyses show that, safely stated, MOX recycling with 1/3 reload fraction can be accommodated for both annual and 18 month fuel cycle schemes in Korean PWRs, without major design modifications on the reactor systems. (author). 12 refs, 4 figs, 3 tabs

  8. Vibration behavior of PWR reactor internals Model experiments and analysis

    International Nuclear Information System (INIS)

    Assedo, R.; Dubourg, M.; Livolant, M.; Epstein, A.

    1975-01-01

    In the late 1971, the CEA and FRAMATOME decided to undertake a comprehensive joint program of studying the vibration behavior of PWR internals of the 900 MWe, 50 cycle, 3 loop reactor series being built by FRAMATOME in France. The PWR reactor internals are submitted to several sources of excitation during normal operation. Two main sources of excitation may effect the internals behavior: the large flow turbulences which could generate various instabilities such as: vortex shedding: the pump pressure fluctuations which could generate acoustic noise in the circuit at frequencies corresponding to shaft speed frequencies or blade passing frequencies, and their respective harmonics. The flow induced vibrations are of complex nature and the approach selected, for this comprehensive program, is semi-empirical and based on both theoretical analysis and experiments on a reduced scale model and full scale internals. The experimental support of this program consists of: the SAFRAN test loop which consists of an hydroelastic similitude of a 1/8 scale model of a PWR; harmonic vibration tests in air performed on full scale reactor internals in the manufacturing shop; the GENNEVILLIERS facilities which is a full flow test facility of primary pump; the measurements carried out during start up on the Tihange reactor. This program will be completed in April 1975. The results of this program, the originality of which consists of studying separately the effects of random excitations and acoustic noises, on the internals behavior, and by establishing a comparison between experiments and analysis, will bring a major contribution for explaining the complex vibration phenomena occurring in a PWR

  9. Feasibility of using gadolinium as a burnable poison in PWR cores. Final report

    International Nuclear Information System (INIS)

    Rothleder, B.M.

    1981-02-01

    As an alternative to the use of lumped burnable absorbers in PWR cores, distributed burnable absorbers are being considered for generic application. These burnable absorbers take the form of Gd 2 O 3 mixed with UO 2 in selected fuel rods (as is currently done in BWR cores). The work discussed herein concerns a three-dimensional feasibility study of the use of such distributed burnable absorbers in PWR cores. This study of distributed burnable absorbers was performed for the first cycle of a typical current design PWR using the following steps: analysis of a generic reference core design; determination of gadolinium assembly designs; determination of a generic gadolinium core design; evaluation of feasibility by examining selected parameters; and redesign of the generic gadolinium core, using axial zoning

  10. PWR thermocouple mechanical sealing structure

    International Nuclear Information System (INIS)

    Shen Qiuping; He Youguang

    1991-08-01

    The PWR in-core temperature detection device, which is one of measures to insure reactor safety operation, is to monitor and diagnose reactor thermal power output and in-core power distribution. The temperature detection device system uses thermocouples as measuring elements with stainless steel protecting sleeves. The thermocouple has a limited service time and should be replaced after its service time has reached. A new sealing device for the thermocouples of reactor in-core temperature detection system has been developed to facilitate replacement. The structure is complete tight under high temperature and pressure without any leakage and seepage, and easy to be assembled or disassembled in radioactive environment. The device is designed to make it possible to replace the thermocouple one by one if necessary. This is a new, simple and practical structure

  11. PWR standardization: The French experience

    International Nuclear Information System (INIS)

    Bacher, P.E.

    1987-01-01

    After a short historical review of the French PWR programme with 45000 MWe in operation and 15000 MWe under construction, the paper first develops the objectives and limits of the standardizatoin policy. Implementation of standardization is described through successive reactor series and feedback of experience, together with its impact on safety and on codes and standards. Present benefits of standardization range from low engineering costs to low backfitting costs, via higher quality, reduction in construction times and start-up schedules and improved training of operators. The future of the French programme into the 1990's is again with an advanced standardized series, the N4-1400 MW plant. There is no doubt that the very positive experience with standardization is relevant to any country trying to achieve self-reliance in the nuclear power field. (author)

  12. Maintenance technologies for SCC of PWR

    International Nuclear Information System (INIS)

    Okimura, Koji; Hori, Nobuyuki; Kanzaki, Hiroshi; Tokuhisa, Kiichi; Kamo, Kazuhiko; Kurokawa, Masaaki

    2007-01-01

    The recent technologies of test, relaxation of deterioration, repairing and change of materials are explained for safe and stable operation of pressurized water reactor (PWR). Stress corrosion cracking (SCC) is originated by three factors such as materials, stress and environment. The eddy current test (ECT) method for the stream generator pipe and the ultrasonic test method for welding part of pipe were developed as the test technologies. Primary water stress corrosion cracking (PWSCC) of Inconel 600 in the welding part is explained. The shot peening of instrument in the gas, the water jet peening of it in water, and laser irradiation on the surface are illustrated as some examples of improvement technology of stress. The cladding of Inconel 690 on Inconel 600 is carried out under the condition of environmental cut. Total or some parts of the upper part of reactor, stream generator and structure in the reactor are changed by the improvement technologies. Changing Inconel 600 joint in the exit pipe of reactor with Inconel 690 is illustrated. (S.Y.)

  13. Results of water chemistry control in the in-pile ''Callisto'' loop (an experimental PWR rig installed in the BR2 reactor)

    International Nuclear Information System (INIS)

    Weber, M.; Benoit, P.; Dekeyser, J.; Verwimp, A.

    1994-01-01

    Since June 1992, a new experimental facility, called CALLISTO, is being irradiated in the BR2 materials testing reactor at Mol, Belgium. The main objective of the present test campaign is to study the behaviour of advanced fuel to high burn-up rates in a realistic PWR environment. Three in-pile sections, containing each 9 fuel rods, are loaded inside the reactor vessel and are connected to a common out-of-pile pressurized water circulation loop (ref.1). The later is branched-off into a purification circuit (feed-bleed concept) and further equipped with safety and auxiliary systems. To cope with the test programme, the equipments are designed so that the guidelines of a PWR primary water chemistry can be followed (ref.2). Real steady-state conditions cannot be observed because the typical BR2 cycle (3 weeks running/3 weeks shut-down) is much shorter and because the rig is cooled down during each reactor shut-down. The purpose of this poster is to provide results of chemical parameters recorded during the cycling behaviour of the CALLISTO primary water. (authors). 4 figs., 1 tab., 2 refs

  14. Model for calculating the boron concentration in PWR type reactors

    International Nuclear Information System (INIS)

    Reis Martins Junior, L.L. dos; Vanni, E.A.

    1986-01-01

    A PWR boron concentration model has been developed for use with RETRAN code. The concentration model calculates the boron mass balance in the primary circuit as the injected boron mixes and is transported through the same circuit. RETRAN control blocks are used to calculate the boron concentration in fluid volumes during steady-state and transient conditions. The boron reactivity worth is obtained from the core concentration and used in RETRAN point kinetics model. A FSAR type analysis of a Steam Line Break Accident in Angra I plant was selected to test the model and the results obtained indicate a sucessfull performance. (Author) [pt

  15. Estimating probable flaw distributions in PWR steam generator tubes

    International Nuclear Information System (INIS)

    Gorman, J.A.; Turner, A.P.L.

    1997-01-01

    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

  16. The N4 plant: culmination of French PWR experience

    International Nuclear Information System (INIS)

    Bellet, J.; Houyez, A.; Journet, J.; Pierrard, J.H.

    1985-01-01

    The model N4 series of 1400MWe class PWR plants has been developed in France from a unique base of technical and operating experience. It meets the French government's requirement for a reactor free of constraints due to licensing agreements with overseas companies, with enhanced safety features and incorporating the lessons of Three Mile Island. In particular, improvements have been made to the reactor vessel, the steam generators, the primary pumps and control systems. The units are capable of daily load following and extended operation between refuelling. The N4 plant includes a new design of turbine-generator. (author)

  17. Babcock and Wilcox advanced PWR development

    International Nuclear Information System (INIS)

    Kulynych, G.E.; Lemon, J.E.

    1986-01-01

    The Babcock and Wilcox 600 MWe PWR design is discussed. Main features of the new B-600 design are improvements in reactor system configuration, glandless coolant pumps, safety features, core design and steam generators

  18. Hydrogen production in a PWR during LOCA

    International Nuclear Information System (INIS)

    Cassette, P.

    1983-12-01

    The purpose of this paper is to provide information on hydrogen generation during LOCA in French 900 MW PWR power plants. The design basis accident is taken into account as well as more severe accidents assuming failure of emergency systems

  19. Recommendations of the MRP-139: Inspection of Welds dissimilar in Nozzles PWR reactor vessel in Spain

    International Nuclear Information System (INIS)

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

    2010-01-01

    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.

  20. Experiments on natural circulation during PWR severe accidents and their analysis

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Stewart, W.A.; Sha, W.T.

    1988-01-01

    Buoyancy-induced natural circulation flows will occur during the early-part of PWR high pressure accident scenarios. These flows affect several key parameters; in particular, the course of such accidents will most probably change due to local failures occurring in the primary coolant system (CS) before substantial core degradation. Natural circulation flow patterns were measured in a one-seventh scale PWR PCS facility at Westinghouse RandD laboratories. The measured flow and temperature distributions are report in this paper. The experiments were analyzed with the COMMIX code and good agreement was obtained between data and calculations. 10 refs., 8 figs., 2 tabs

  1. Application on electrochemistry measurement of high temperature high pressure condition in PWR nuclear power plants

    International Nuclear Information System (INIS)

    Li Yuchun; Xiao Zhongliang; Jiang Ya; Yu Xiaowei; Pang Feifei; Deng Fenfang; Gao Fan; Zhou Nianguang

    2011-01-01

    High temperature high pressure electrochemistry testing system was comprehensively analyzed in this paper, according to actual status for supervision in primary and secondary circuits of PWR nuclear power plants. Three research methods were reviewed and discussed for in-situ monitor system. By combination with ECP realtime measurement it was executed for evaluation and water chemistry optimization in nuclear power plants. It is pointed out that in-situ electrochemistry measurement has great potential application for water chemistry evaluation in PWR nuclear power plants. (authors)

  2. Study on quality control measures of static casting main pipe in PWR nuclear power plant

    International Nuclear Information System (INIS)

    Jiang Zhenbiao; Li Guanying; Liu Zhicheng

    2013-01-01

    This study analyzes the main reasons which impact the quality of primary pipe static casting elbows in PWR-M310 nuclear power plant. The quality control measures are developed from the election and inspection of material, improving sand production and casting process, improving lean management of personnel. The static casting defects of primary pipe elbows for Fuqing Unit 1 and 2 were down to less than 50% of the former project. The quality of static casting for the primary pipe elbows was significantly improved. Moreover, the implementation saves human resources and financing to repair casting defects, and also helps to win the delivery schedule. The quality control measures are good reference for improving primary pipe casting process. This study provides valuable experience for further study of improving the quality of static casting for the primary pipe of PWR nuclear power plant. (authors)

  3. Activity incorporation into zinc doped PWR oxides

    International Nuclear Information System (INIS)

    Maekelae, Kari

    1998-01-01

    Activity incorporation into the oxide layers of PWR primary circuit constructional materials has been studied in Halden since 1993. The first zinc injection tests showed that zinc addition resulted in thinner oxide layers on new metal surfaces and reduced further incorporation of activity into already existing oxides. These tests were continued to find out the effects of previous zinc additions on the pickup of activity onto the surface oxides which were subsequently exposed to zinc-free coolant. The results showed that previous zinc addition will continue to reduce the rate of Co-60 build-up on out-of-core surfaces in subsequent exposure to zinc-free coolants. However, the previous Zn free test was performed for a relatively short period of time and the water chemistry programme was continued to find out the long term effects for extended periods without zinc. The activity incorporation into the stainless steel oxides started to increase as soon as zinc dosing to the coolant was stopped. The Co-60 concentration was lowest on all of the coupons which were first oxidised in Zn containing primary coolant. After the zinc injection period the thickness of the oxides increased, but activity in the oxide films did not increase at the same rate. This could indicate that zinc in the oxide blocks the adsorption sites for Co-60 incorporation. The Co-60 incorporation rate into the oxides on Inconel 600 seemed to be linear whether the oxide was pre-oxidised with or without Zn. The results indicate that zinc can either replace or prevent cobalt transport in the oxides. The results show that for zinc injection to be effective it should be carried out continuously. Furthermore the actual mechanism by which Zn inhibits the activity incorporation into the oxides is still not clear. Therefore, additional work has to follow with specified materials to verify the conclusions drawn in this work. (author)

  4. Integral type small PWR with stand-alone safety

    International Nuclear Information System (INIS)

    Makihara, Yoshiaki

    2001-01-01

    A feasibility study is achieved on an integral type small PWR with stand-alone safety. It is designed to have the following features. (1) The coolant does not leak out at any accidental condition. (2) The fuel failure does never occur while it is supposed on the large scale PWR at the design base accident. (3) At any accidental condition the safety is secured without any support from the outside (stand-alone safety secure). (4) It has self-regulating characteristics and easy controllability. The above features can be satisfied by integrate the steam generator and CRDM in the reactor vessel while the pipe line break has to be considered on the conventional PWR. Several counter measures are planned to satisfy the above features. The economy feature is also attained by several simplifications such as (1) elimination of main coolant piping and pressurizer by the integration of primary cooling system and self-pressurizing, (2) elimination of RCP by application of natural circulating system, (3) elimination of ECCS and accumulator by application of static safety system, (4) large scale volume reduction of the container vessel by application of integrated primary cooling system, (5) elimination of boric acid treatment by deletion of chemical shim. The long operation period such as 10 years can be attained by the application of Gd fuel in one batch refueling. The construction period can be shortened by the standardizing the design and the introduction of modular component system. Furthermore the applicability of the reduced modulation core is also considered. (K. Tsuchihashi)

  5. Reactor analysis support package (RASP). Volume 7. PWR set-point methodology. Final report

    International Nuclear Information System (INIS)

    Temple, S.M.; Robbins, T.R.

    1986-09-01

    This report provides an overview of the basis and methodology requirements for determining Pressurized Water Reactor (PWR) technical specifications related setpoints and focuses on development of the methodology for a reload core. Additionally, the report documents the implementation and typical methods of analysis used by PWR vendors during the 1970's to develop Protection System Trip Limits (or Limiting Safety System Settings) and Limiting Conditions for Operation. The descriptions of the typical setpoint methodologies are provided for Nuclear Steam Supply Systems as designed and supplied by Babcock and Wilcox, Combustion Engineering, and Westinghouse. The description of the methods of analysis includes the discussion of the computer codes used in the setpoint methodology. Next, the report addresses the treatment of calculational and measurement uncertainties based on the extent to which such information was available for each of the three types of PWR. Finally, the major features of the setpoint methodologies are compared, and the principal effects of each particular methodology on plant operation are summarized for each of the three types of PWR

  6. PWR vessel flaw distribution development

    International Nuclear Information System (INIS)

    Rosinski, S.T.; Kennedy, E.L.; Foulds, J.R.; Kinsman, K.M.

    1990-01-01

    This paper reports on PWR pressure vessels which operate under NRC rules and regulatory guides intended to prevent failure of the vessels. Plants failing to meet the operating criteria specified under these rules and regulations are required to analytically demonstrate fitness for service in order to continue operation. The initial flaw size or distribution of initial vessel flaws is a key input to the required vessel integrity analyses. However, the flaw distribution assumed in the development of the NRC Regulations and recommended for the plant specific analyses is potentially over-conservative. This is because the distribution is based on the limited amount of vessel inspection data available at the time the criteria were being developed and does not take full advantage of the more recent and reliable domestic vessel inspection results. The U.S. Department of Energy is funding an effort through Sandia National Laboratories to investigate the possibility of developing a new flaw distribution based on the increased amount and improved reliability of domestic vessel inspection data. Results of Phase I of the program indicate that state-of-the-art NDE systems' capabilities are sufficient for development of a new flaw distribution that could ultimately provide life extension benefits over the presently required operating practice

  7. Upgrading of PWR plant simulators

    International Nuclear Information System (INIS)

    Wada, Tomonori; Sasaki, Kazunori; Nakaishi, Hirokazu.

    1989-01-01

    For the education and training of operators in electric power plants, simulators have been employed, and it is well known that their effect is great. There are operation training simulators which simulate the dynamic characteristics of plants and all the machinery and equipment that operators handle, and train the procedure of restoration at the time of abnormality in plants, education simulators which can analyze the dynamic characteristics of plants efficiently in a short time, and offer information by visualizing phenomena with three-dimensional display and others so as to be easily understandable, and forecast simulators which do the analysis forecasting plant behavior at the time of abnormality in plants, and investigate the necessity of the guide for operation procedure and the countermeasures at the time of emergency. In this explanation, the upgrading of operation training simulators which have been put already in training is discussed. The constitution of simulator system and the instructor function, the outline of PWR plant simulation models comprising thermal flow model, pump model, leak model and so on, the techniques of increasing simulator speed, and the example of analysis using the NUPAC code are reported. (K.I.)

  8. PWR secondary water chemistry study

    International Nuclear Information System (INIS)

    Pearl, W.L.; Sawochka, S.G.

    1977-02-01

    Several types of corrosion damage are currently chronic problems in PWR recirculating steam generators. One probable cause of damage is a local high concentration of an aggressive chemical even though only trace levels are present in feedwater. A wide variety of trace chemicals can find their way into feedwater, depending on the sources of condenser cooling water and the specific feedwater treatment. In February 1975, Nuclear Water and Waste Technology Corporation (NWT), was contracted to characterize secondary system water chemistry at five operating PWRs. Plants were selected to allow effects of cooling water chemistry and operating history on steam generator corrosion to be evaluated. Calvert Cliffs 1, Prairie Island 1 and 2, Surry 2, and Turkey Point 4 were monitored during the program. Results to date in the following areas are summarized: (1) plant chemistry variations during normal operation, transients, and shutdowns; (2) effects of condenser leakage on steam generator chemistry; (3) corrosion product transport during all phases of operation; (4) analytical prediction of chemistry in local areas from bulk water chemistry measurements; and (5) correlation of corrosion damage to chemistry variation

  9. Is our Universe typical?

    International Nuclear Information System (INIS)

    Gurzadyan, V.G.

    1988-01-01

    The problem of typicalness of the Universe - as a dynamical system possessing both regular and chaotic regions of positive measure of phase space, is raised and discussed. Two dynamical systems are considered: 1) The observed Universe as a hierarchy of systems of N graviting bodies; 2) (3+1)-manifold with matter evolving to Wheeler-DeWitt equation in superspace with Hawking boundary condition of compact metrics. It is shown that the observed Universe is typical. There is no unambiguous answer for the second system yet. If it is typical too then the same present state of the Universe could have been originated from an infinite number of different initial conditions the restoration of which is practically impossible at present. 35 refs.; 2 refs

  10. 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)

  11. Typical Complexity Numbers

    Indian Academy of Sciences (India)

    First page Back Continue Last page Overview Graphics. Typical Complexity Numbers. Say. 1000 tones,; 100 Users,; Transmission every 10 msec. Full Crosstalk cancellation would require. Full cancellation requires a matrix multiplication of order 100*100 for all the tones. 1000*100*100*100 operations every second for the ...

  12. Liquid radioactive waste processing improvement of PWR nuclear power plants

    International Nuclear Information System (INIS)

    Nery, Renata Wolter dos Reis; Martinez, Aquilino Senra; Monteiro, Jose Luiz Fontes

    2005-01-01

    The study evaluate an inorganic ion exchange to process the low level liquid radwaste of PWR nuclear plants, so that the level of the radioactivity in the effluents and the solid waste produced during the treatment of these liquid radwaste can be reduced. The work compares two types of ion exchange materials, a strong acid cation exchange resin, that is the material typically used to remove radionuclides from PWR nuclear plants wastes, and a mordenite zeolite. These exchange material were used to remove cesium from a synthetic effluent containing only this ion and another effluent containing cesium and cobalt. The breakthrough curves of the zeolite and resin using a fix bed reactor were compared. The results demonstrated that the zeolite is more efficient than the resin in removing cesium from a solution containing cesium and cobalt. The results also showed that a bed combining zeolite and resin can process more volume of an effluent containing cesium and cobalt than a bed resin alone. (author)

  13. Parallel GPU implementation of PWR reactor burnup

    International Nuclear Information System (INIS)

    Heimlich, A.; Silva, F.C.; Martinez, A.S.

    2016-01-01

    Highlights: • Three GPU algorithms used to evaluate the burn-up in a PWR reactor. • Exhibit speed improvement exceeding 200 times over the sequential. • The C++ container is expansible to accept new nuclides chains. - Abstract: This paper surveys three methods, implemented for multi-core CPU and graphic processor unit (GPU), to evaluate the fuel burn-up in a pressurized light water nuclear reactor (PWR) using the solutions of a large system of coupled ordinary differential equations. The reactor physics simulation of a PWR reactor spends a long execution time with burnup calculations, so performance improvement using GPU can imply in better core design and thus extended fuel life cycle. The results of this study exhibit speed improvement exceeding 200 times over the sequential solver, within 1% accuracy.

  14. ABB advanced BWR and PWR fuel

    International Nuclear Information System (INIS)

    Junkrans, S.; Helmersson, S.; Andersson, S.

    1999-01-01

    Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both BWR and PWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter, proven to meet the -6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10x10 BWR fuel, where ABB is the only vendor to date with multi batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of BWR and PWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its customers. (orig.)

  15. Preliminary study on direct recycling of spent PWR fuel in PWR system

    International Nuclear Information System (INIS)

    Waris, Abdul; Nuha; Novitriana; Kurniadi, Rizal; Su'ud, Zaki

    2012-01-01

    Preliminary study on direct recycling of PWR spent fuel to support SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario has been conducted. Several spent PWR fuel compositions in loaded PWR fuel has been evaluated to obtain the criticality of reactor. The reactor can achieve it criticality for U-235 enrichment in the loaded fresh fuel is at least 4.0 a% with the minimum fraction of the spent fuel in the core is 15.0 %. The neutron spectra become harder with the escalating of U-235 enrichment in the loaded fresh fuel as well as the amount of the spent fuel in the core.

  16. Typicality and reasoning fallacies.

    Science.gov (United States)

    Shafir, E B; Smith, E E; Osherson, D N

    1990-05-01

    The work of Tversky and Kahneman on intuitive probability judgment leads to the following prediction: The judged probability that an instance belongs to a category is an increasing function of the typicality of the instance in the category. To test this prediction, subjects in Experiment 1 read a description of a person (e.g., "Linda is 31, bright, ... outspoken") followed by a category. Some subjects rated how typical the person was of the category, while others rated the probability that the person belonged to that category. For categories like bank teller and feminist bank teller: (1) subjects rated the person as more typical of the conjunctive category (a conjunction effect); (2) subjects rated it more probable that the person belonged to the conjunctive category (a conjunction fallacy); and (3) the magnitudes of the conjunction effect and fallacy were highly correlated. Experiment 2 documents an inclusion fallacy, wherein subjects judge, for example, "All bank tellers are conservative" to be more probable than "All feminist bank tellers are conservative." In Experiment 3, results parallel to those of Experiment 1 were obtained with respect to the inclusion fallacy.

  17. Typicals/Típicos

    Directory of Open Access Journals (Sweden)

    Silvia Vélez

    2004-01-01

    Full Text Available Typicals is a series of 12 colour photographs digitally created from photojournalistic images from Colombia combined with "typical" craft textiles and text from guest writers. Typicals was first exhibited as photographs 50cm x 75cm in size, each with their own magnifying glass, at the Contemporary Art Space at Gorman House in Canberra, Australia, in 2000. It was then exhibited in "Feedback: Art Social Consciousness and Resistance" at Monash University Museum of Art in Melbourne, Australia, from March to May 2003. From May to June 2003 it was exhibited at the Museo de Arte de la Universidad Nacional de Colombia Santa Fé Bogotá, Colombia. In its current manifestation the artwork has been adapted from the catalogue of the museum exhibitions. It is broken up into eight pieces corresponding to the contributions of the writers. The introduction by Sylvia Vélez is the PDF file accessible via a link below this abstract. The other seven PDF files are accessible via the 'Supplementary Files' section to the left of your screen. Please note that these files are around 4 megabytes each, so it may be difficult to access them from a dial-up connection.

  18. PWR cold-leg small break loca with faulty HPI

    International Nuclear Information System (INIS)

    Kumamaru, H.; Kukita, Y.

    1991-01-01

    The ROSA-IV Large Scale Test Facility (LSTF) is a 1/48 volumetrically-scaled model of a pressurized water reactor (PWR). At the LSTF are performed cold-leg small-break loss-of-coolant accident (LOCA) tests with faulty high pressure injection (HPI) system for break areas from 0.5% to 10% and an intentional primary system depressurization test following a small-break LOCA test. A simple prediction model is proposed for prediction of times of major events. Test data and calculations show that intentional primary system depressurization with use of the pressurizer power-operated relief valves (PORVs) is effective for break areas of approximately 0.5% or less, is unnecessary for breaks of 5% or more, and is insufficient for intermediate break areas to maintain adequate core cooling. (author)

  19. Study of PWR reactor efficiency as a function of turbine steam extractions; Estudo da otimizacao da eficiencia de reator PWR em funcao das extracoes de vapor da turbina

    Energy Technology Data Exchange (ETDEWEB)

    Rocha, Janine Gandolpho da; Alvim, Antonio Carlos Marques; Martinez, Aquilino Senra [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear

    2002-07-01

    The objective of this work is to optimize the extractions of the low-pressure turbine of a PWR nuclear reactor, in order to obtain the best thermodynamic cycle efficiency. We have analyzed typical data of a 1300 MW PWR reactor, operating at 25%, 50%, 75% and 100% capacities, respectively. The first stage of this study consists of generating a mathematical model capable of describing the reactor behavior and efficiency at any power level. The second stage of this study consists of to combine the generated mathematical model in an optimization computer program that optimize the extractions flow of the low-pressure turbine until it finds the optimal system efficiency. This work does not alter the nuclear facility project in any way. (author)

  20. Benchmark calculations on resonance absorption by 238U in a PWR pin-cell geometry

    International Nuclear Information System (INIS)

    Kruijf, W.J.M. de; Janssen, A.J.

    1993-12-01

    Very accurate Monte Carlo calculations with MCNP have been performed to serve as a reference for benchmark calculations on resonance absorption by 238 U in a typical PWR pin-cell geometry. Calculations with the energy-pointwise slowing down code ROLAIDS-CPM show that this code calculates the resonance absorption accurately. Calculations with the multigroup discrete ordinates code XSDRN show that accurate results can only be achieved with a very fine energy mesh. (orig.)

  1. 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.

  2. Thermohidraulic model for a typical steam generator of PWR Nuclear Power Plants

    International Nuclear Information System (INIS)

    Braga, C.V.M.

    1980-06-01

    A model of thermohidraulic simulation, for steady state, considering the secondary flow divided in two parts individually homogeneous, and with heat and mass transferences between them is developed. The quality of the two-phase mixture that is fed to the turbine is fixed and, based on this value, the feedwater pressure is determined. The recirculation ratio is intrinsically determined. Based on this model it was developed the GEVAP code, in Fortran-IV language. The model is applied to the steam generator of the Angra II nuclear power plant and the results are compared with KWU'S design parameters, being considered satisfactory. (Author) [pt

  3. Vulnerability analysis of a PWR to an external event

    International Nuclear Information System (INIS)

    Aruety, S.; Ilberg, D.; Hertz, Y.

    1980-01-01

    The Vulnerability of a Nuclear Power Plant (NPP) to external events is affected by several factors such as: the degree of redundancy of the reactor systems, subsystems and components; the separation of systems provided in the general layout; the extent of the vulnerable area, i.e., the area which upon being affected by an external event will result in system failure; and the time required to repair or replace the systems, when allowed. The present study offers a methodology, using Probabilistic Safety Analysis, to evaluate the relative importance of the above parameters in reducing the vulnerability of reactor safety systems. Several safety systems of typical PWR's are analyzed as examples. It was found that the degree of redundancy and physical separation of the systems has the most prominent effect on the vulnerability of the NPP

  4. Dose rates modeling of pressurized water reactor primary loop components with SCALE6.0

    International Nuclear Information System (INIS)

    Matijević, Mario; Pevec, Dubravko; Trontl, Krešimir

    2015-01-01

    Highlights: • Shielding analysis of the typical PWR primary loop components was performed. • FW-CADIS methodology was thoroughly investigated using SCALE6.0 code package. • Versatile ability of SCALE6.0/FW-CADIS for deep penetration models was proved. • The adjoint source with focus on specific material can improve MC modeling. - Abstract: The SCALE6.0 simulation model of a typical PWR primary loop components for effective dose rates calculation based on hybrid deterministic–stochastic methodology was created. The criticality sequence CSAS6/KENO-VI of the SCALE6.0 code package, which includes KENO-VI Monte Carlo code, was used for criticality calculations, while neutron and gamma dose rates distributions were determined by MAVRIC/Monaco shielding sequence. A detailed model of a combinatorial geometry, materials and characteristics of a generic two loop PWR facility is based on best available input data. The sources of ionizing radiation in PWR primary loop components included neutrons and photons originating from critical core and photons from activated coolant in two primary loops. Detailed calculations of the reactor pressure vessel and the upper reactor head have been performed. The efficiency of particle transport for obtaining global Monte Carlo dose rates was further examined and quantified with a flexible adjoint source positioning in phase-space. It was demonstrated that generation of an accurate importance map (VR parameters) is a paramount step which enabled obtaining Monaco dose rates with fairly uniform uncertainties. Computer memory consumption by the S N part of hybrid methodology represents main obstacle when using meshes with large number of cells together with high S N /P N parameters. Detailed voxelization (homogenization) process in Denovo together with high S N /P N parameters is essential for precise VR parameters generation which will result in optimized MC distributions. Shielding calculations were also performed for the reduced PWR

  5. A Feasibility Study on Core Cooling of Reduced-Moderation PWR for the Large Break LOCA

    International Nuclear Information System (INIS)

    Hiroyuki Yoshida; Akira Ohnuki; Hajime Akimoto

    2002-01-01

    A design study of a reduced-moderation water reactor (RMWR) with tight lattice core is being carried out at the Japan Atomic Energy Research Institute (JAERI) as one candidate for future reactors. The concept is developed to achieve a conversion ratio greater than unity using the tight lattice core (volume ratio of moderator to fuel is around 0.5 and the gap spacing between the fuel rods is remarkably narrower than in a reactor currently operated). Under such tight configuration, the core thermal margin becomes smaller and should be evaluated in a normal operation and also during the reflood phase in a large break loss-of-coolant accident (LBLOCA) for PWR type reactors. In this study, we have performed a feasibility evaluation on core cooling of reduced moderation PWR for the LBLOCA (200% break). The evaluation was performed for the primary system after the break by the REFLA/TRAC code. The core thermal output of the reduced moderation PWR is 2900 MWt, the gap between adjacent fuel rods is 1 mm, and heavy water is used as the moderator and coolant. The present design adopts seed fuel assemblies (MOX fuel) and several blanket fuel assemblies. In the blanket fuel assemblies, power density is lower than that of the seed fuel assemblies. Then, we set a channel box to each fuel assembly in order to adjust the flow rate in each assembly, because the possibility that the coolant boils in the seed fuel assemblies is very high. The pressure vessel diameter is bigger in comparison with a current PWR and core height is smaller than the current one. The current 4-loop PWR system is used, and, however, to fit into the bigger pressure vessel volume (about 1.5 times), we set up the capacity of the accumulator (1.5 times of the current PWR). Although the maximum clad temperature reached at about 1200 K in the position of 0.6 m from the lower core support plate, it is sufficiently lower than the design criteria of the current PWR (1500 K). The core cooling of the reduced moderation

  6. PWR reactors for BBR nuclear power plants

    International Nuclear Information System (INIS)

    Structure and functioning of the nuclear steam generator system developed by BBR and its components are described. Auxiliary systems, control and load following behaviour and fuel management are discussed and the main data of PWR given. The brochure closes with a perspective of the future of the Muelheim-Kaerlich nuclear power plant. (GL) [de

  7. Reliability of PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Ribeiro, A.A.T.; Muniz, A.A.

    1978-12-01

    Results of the analysis of factors influencing the reliability of international nuclear power plants of the PWR type are presented. The reliability factor is estimated and the probability of its having lower values than a certain specified value is discussed. (Author) [pt

  8. Coolant monitoring systems for PWR reactors

    International Nuclear Information System (INIS)

    Luzhnov, A.M.; Morozov, V.V.; Tsypin, S.G.

    1987-01-01

    The ways of improving information capacity of existing monitoring systems and the necessity of designing new ones for coolant monitoring are reviewed. A wide research program on development of coolant monitoring systems in PWR reactors is analyzed. The possible applications of in-core and out-of-core detectors for coolant monitoring are demonstrated

  9. Secondary systems of PWR and BWR

    International Nuclear Information System (INIS)

    Schindler, N.

    1981-01-01

    The secondary systems of a nuclear power plant comprises the steam, condensate and feedwater cycle, the steam plant auxiliary or ancillary systems and the cooling water systems. The presentation gives a general review about the main systems which show a high similarity of PWR and BWR plants. (orig./RW)

  10. Simulation model of a PWR power plant

    International Nuclear Information System (INIS)

    Larsen, N.

    1987-03-01

    A simulation model of a hypothetical PWR power plant is described. A large number of disturbances and failures in plant function can be simulated. The model is written as seven modules to the modular simulation system for continuous processes DYSIM and serves also as a user example of this system. The model runs in Fortran 77 on the IBM-PC-AT. (author)

  11. Utilization of thorium in PWR type reactors

    International Nuclear Information System (INIS)

    Correa, F.

    1977-01-01

    Uranium 235 consumption is comparatively evaluated with thorium cycle for a PWR type reactor. Modifications are only made in fuels components. U-235 consumption is pratically unchanged in both cycles. Some good results are promised to the mixed U-238/Th-232 fuel cycle in 1/1 proportion [pt

  12. Status of developing advanced PWR in Japan

    International Nuclear Information System (INIS)

    Iida, Yotaro

    1982-01-01

    During past eleven years since the first PWR power plant, Mihama Unit 1 of Kansai Electric Power Co., started the commercial operation in 1970, Mitsubishi Heavy Industries has endeavored to improve PWR technologies on the basis of the advice from electric power companies and the technical information to overcome difficulties in PWR power plants. Now, the main objective is to improve the overall plant performance, and the rate of operation of Japanese PWR power plants has significantly risen. The improvement of the reliability, the shortening of regular inspection period and the reduction of radioactive waste handling were attempted. In view of the satisfactory operational experience of Westinghouse type PWRs, the basic reactor concept has not been changed so far. Mitsubishi and Westinghouse reached basic agreement in August, 1981, to develop a spectral shift type large capacity reactor as the advanced PWRs for Japan. This type of PWRs hab higher degree of freedom for extended fuel cycle operation and enhances the advantage of entire fuel cycle economy, particularly the significant reduction of uranium use. The improved neutron economy is attainable by reducing neutron loss, and the core design with low power density and the economical use of plutonium are advantageous for the fuel cycle economy. (Kako, I.)

  13. An evaluation of tight - pitch PWR cores

    International Nuclear Information System (INIS)

    Correa, F.

    1980-01-01

    The subtask of a project carried out at MIT (Massachusetts Institute of Technology) for DOE (Department of Energy) as part of their NASAP/INFCE - related effects involving the optimization of PWR lattices in the recycle model is summarized. (E.G.) [pt

  14. Survey of PWR water chemistry

    International Nuclear Information System (INIS)

    Gorman, J.

    1989-02-01

    This report surveys available information regarding primary and secondary water chemistries of pressurized water reactors (PWRs) and the impact of these water chemistries on reactor operation. The emphasis of the document is on aspects of water chemistry that affect the integrity of the primary pressure boundary and the radiation dose associated with maintenance and operation. The report provides an historical overview of the development of primary and secondary water chemistries, and describes practices currently being followed. Current problems and areas of research associated with water chemistry are described. Recommendations for further research are included. 183 refs., 9 figs., 19 tabs

  15. Improvements to PWR type reactors

    International Nuclear Information System (INIS)

    Ailloud, Jean; Monteil, Marcel.

    1978-01-01

    Improvements to pressurized water nuclear reactors are described, where the core coolant, called primary fluid, flows under the effect of a circulating pump in a primary loop between a steam generator and a pressure vessel containing the reactor core. The steam generator includes a bundle of tubes through which flows the primary fluid which exchanges calories with a secondary fluid, generally water, entering the generator as a liquid and issuing from it as steam. After expansion in turbines and recovery in a condenser, this steam is returned to the inside of the generator. Each primary fluid circulating pump is powered by a back-pressure turbine located in parallel with the high pressure section of the main turbine and hence fed with steam taken directly from the steam generator or the main steam pipe outside it [fr

  16. Study on thermo-hydraulic behavior during reflood phase of a PWR-LOCA

    International Nuclear Information System (INIS)

    Sugimoto, Jun

    1989-01-01

    This paper describes thermo-hydraulic behavior during the reflood phase in a postulated large-break loss-of-coolant accident (LOCA) of a PWR. In order to better predict the reflood transient in a nuclear safety analysis specific analytical models have been developed for, saturated film boiling heat transfer in inverted slung flow, the effect of grid spacers on core thermo-hydraulics, overall system thermo-hydraulic behavior, and the thermal response similarity between nuclear fuel rods and simulated rods. A heat transfer correlation has been newly developed for saturated film boiling based on a 4 x 4-rod experiment conducted at JAERI. The correlation provides a good agreement with existing experiments except in the vicinity of grid spacer locations. An analytical model has then been developed addressing the effect of grid spacers. The thermo-hydraulic behavior near the grid spacers was found to be predicted well with this model by considering the breakup of droplets in dispersed flow and water accumulation above the grid spacers in inverted slung flow. A system analysis code has been developed which couples the one-dimensional core and multi-loop primary system component models. It provides fairly good agreement with system behavior obtained in a large-scale integral reflood experiment with active primary system components. An analytical model for the radial temperature distribution in a rod has been developed and verified with data from existing experiments. It was found that a nuclear fuel rod has a lower cladding temperature and an earlier quench time than an electrically heated rod in a typical reflood condition. (author)

  17. LOFT/LP-FW-1, Loss of Fluid Test, PWR Response to Loss-of-Feedwater Transient

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The first OECD LOFT experiment was conducted on February 20, 1983. It was designed to evaluate the generic PWR system response during a complete loss-of-feedwater transient. The objective of the experiment was to investigate the performance of primary 'feed and bleed' using a 'bleed' from the PORV and 'feed' from the HPIS to provide decay heat removal and system pressure reduction while maintaining the primary coolant inventory. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  18. Crack growth rate of PWR piping

    International Nuclear Information System (INIS)

    Bethmont, M.; Doyen, J.J.; Lebey, J.

    1979-01-01

    The Aquitaine 1 program, carried out jointly by FRAMATOME and the CEA is intended to improve knowledge about cracking mechanisms in AISI 316 L austenitic stainless steel under conditions similar to those of the PWR environment (irradiation excluded). Experiments of fatigue crack growth are performed on piping elements, scale 1/4 of primary pipings, by means of internal hydraulic cyclic pressure. Interpretation of results requires a knowledge of the stress intensity factor Ksub(I) at the front of the crack. Results of a series of calculations of Ksub(I) obtained by different methods for defects of finite and infinite length (three dimensional calculations) are given in the paper. The following have been used: calculations by finite elements, calculations by weight function. Notches are machined on the test pipes, which are subjected to internal hydraulic pressure cycles, under cold conditions, to initiate a crack at the tip of the notch. They are then cycled at a frequency of 4 cycles/hour on on water demineralised loop at a temperature of 280 0 C, the pressure varying at each cycle between approximately 160 bars and 3 bars. After each test, a specimen containing the defect is taken from the pipe for micrographic analysis. For the first test the length of the longitudinal external defect is assumed infinite. The number of cycles carried out is 5880 cycles. Two defects are machined in the tube for the second test. The number of cycles carried out is N = 440. The tests are performed under hot conditions (T = 280 0 C). For the third test two defects are analysed under cold and hot conditions. The number of cycles carried out for the external defect is 7000 when hot and 90000 when cold. The number of cycles for the internal defect is 1650 when hot and 68000 when cold. In order to interpret the results, the data da/dN are plotted on a diagram versus ΔK. Comparisons are made between these results and the curves from laboratory tests

  19. Aqueous Boric acid injection facility of PWR type reactor

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi; Iwami, Masao.

    1996-01-01

    If a rupture should be caused in a secondary system of a PWR type reactor, pressure of a primary coolant recycling system is lowered, and a back flow check valve is opened in response to the lowering of the pressure. Then, low temperature aqueous boric acid in the lower portion of a pressurized tank is flown into the primary coolant recycling system based on the pressure difference, and the aqueous boric acid reaches the reactor core together with coolants to suppress reactivity. If the injection is continued, high temperature aqueous boric acid in the upper portion boils under a reduced pressure, further urges the low temperature aqueous boric acid in the lower portion by the steam pressure and injects the same to the primary system. The aqueous boric acid stream from the pressurized tank flowing by self evaporation of the high temperature aqueous boric acid itself is rectified by a rectifying device to prevent occurrence of vortex flow, and the steam is injected in a state of uniform stream. When the pressure in the pressurized tank is lowered, a bypass valve is opened to introduce the high pressure fluid of primary system into the pressurized tank to keep the pressure to a predetermined value. When the pressure in the pressurized tank is elevated to higher than the pressure of the primary system, a back flow check valve is opened, and high pressure aqueous boric acid is flown out of the pressurized tank to keep the pressure to a predetermined value. (N.H.)

  20. A Typical Synergy

    Science.gov (United States)

    van Noort, Thomas; Achten, Peter; Plasmeijer, Rinus

    We present a typical synergy between dynamic types (dynamics) and generalised algebraic datatypes (GADTs). The former provides a clean approach to integrating dynamic typing in a statically typed language. It allows values to be wrapped together with their type in a uniform package, deferring type unification until run time using a pattern match annotated with the desired type. The latter allows for the explicit specification of constructor types, as to enforce their structural validity. In contrast to ADTs, GADTs are heterogeneous structures since each constructor type is implicitly universally quantified. Unfortunately, pattern matching only enforces structural validity and does not provide instantiation information on polymorphic types. Consequently, functions that manipulate such values, such as a type-safe update function, are cumbersome due to boilerplate type representation administration. In this paper we focus on improving such functions by providing a new GADT annotation via a natural synergy with dynamics. We formally define the semantics of the annotation and touch on novel other applications of this technique such as type dispatching and enforcing type equality invariants on GADT values.

  1. VAMCIS, a new measuring channel for continuous monitoring of leak rates inside PWR steam generators

    International Nuclear Information System (INIS)

    Champion, G.; Dubail, A.; Lefevre, F.

    1988-01-01

    In order to assess the primary to secondary leakage, radioactive isotopes, formed in the primary coolant as a result of fission or neutron capture, are usually monitored in the pressurized water reactor (PWR) secondary coolant. Conventional methods mainly based on the detection of 133 Xe, tritium, and 41 Ar are widely used on French Electricite de France (EdF) PWRs. Some years ago, it appeared necessary to improve both leak rate assessments and steam generator tube rupture (SGTR) detection. A volumetric activity measuring channel inside steam (VAMCIS) has been developed for this purpose. The SGTR that occurred at the North Anna PWR has focused the attention of safety authorities on this new measuring channel. It is planned to implement VAMCIS at North Anna in order to check the leak rate variations more accurately

  2. START - glass model of PWR

    International Nuclear Information System (INIS)

    Marn, J.; Ramsak, M.

    1998-01-01

    Recognizing the importance of nuclear engineering in the area of process engineering the University of Maribor, Faculty of Mechanical Engineering has invested in procuring and erecting glass model of pressurized water reactor. This paper deals with description of the model, its capabilities, and plans for its use within nuclear engineering community of Slovenia. The model, made primarily of glass, serves three purposes: educational, professional development and research. As an example, medium break loss of coolant accident is presented in the paper. Temperatures within primary and secondary side, and pressure on primary side of reactor coolant system are followed. The characteristic points are emphasized, and commented.(author)

  3. Electrochemical measurements in PWR steam generators to follow crevice chemistry

    International Nuclear Information System (INIS)

    Feron, D.

    1991-01-01

    In PWR steam generator crevices, the evolution of chemistry is important for the understanding of corrosion phenomena. Electrochemical measurements have been performed in high temperature simulated crevice environments in order to follow hideout processes and remedial actions (on-line addition of boric acid). Reported tests have been conducted with model boilers of AJAX facilities. Eccentric and concentric tube support plate crevices have been instrumented with platinum electrodes. Electrochemical measurements have been collected when model boiler was under nominal conditions (primary temperature: 335 deg C, secondary temperature: 280 deg C). They include Electrochemical Impedance Spectroscopy (EIS) and potential measurements: with EIS, sodium and boric acid hideouts have been detected and followed. Potential measurements have been performed in an attempt to measure crevice PH evolution

  4. Waste management aspects of entire PWR LOOP decontamination

    International Nuclear Information System (INIS)

    Murray, A.P.; Roesmer, J.

    1988-01-01

    The waste management parameters for decontamination of an entire PWR primary circuit have been determined for dilute alkaline-permanganate/citric acid (APCA), LOMI, ozone and cerium acid process variations. APCA processes generate the largest waste volumes; over 140 m 3 (5000 ft 3 ) in some cases. This represents a potential disposal cost of one million dollars. The cation regeneration column makes the greatest contribution to the disposal volume. In contrast, the LOMI process generates approximately half as much waste, but it is expected to contain relatively high metal concentrations (200-800 ppm). The ozone and cerium acid processes product the least waste, usually under 45 m 3 . These waste volume estimates represent considerable fractions of a utility's annual disposal volume. Consequently, improved waste processing technology is required, and several approaches are suggested

  5. Numerical regulation of a test facility of materials for PWR

    International Nuclear Information System (INIS)

    Zauq, M.H.

    1982-02-01

    The installation aims at testing materials used in nuclear power plants; tests consists in simulations of a design basis accident (failure of a primary circuit of a PWR type reactor) for a qualification of these materials. A description of the test installation, of the thermodynamic control, and of the control system is presented. The organisation of the software is then given: description of the sequence chaining monitor, operation, list and function of the programs. The analog information processing is also presented (data transmission). A real-time microcomputer and clock are used for this work. The microprocessor is the 6800 of MOTOROLA. The microcomputer used has been built around the MC 6800; its structure is described. The data acquisition include an analog data acquisition system and a numerical data acquisition system. Laboratory and on-site tests are finally presented [fr

  6. PWR-blowdown heat transfer separate effects program

    International Nuclear Information System (INIS)

    Thomas, D.G.

    1976-01-01

    The ORNL Pressurized-Water Reactor Blowdown Heat Transfer (PWR-BDHT) Program is an experimental separate-effects study of the relations among the principal variables that can alter the rate of blowdown, the presence of flow reversal and rereversal, time delay to critical heat flux, the rate at which dryout progresses, and similar time-related functions that are important to LOCA analysis. Primary test results are obtained from the Thermal-Hydraulic Test Facility (THTF). Supporting experiments are carried out in several additional test loops - the Forced Convection Test Facility (FCTF), an air-water loop, a transient steam-water loop, and a low-temperature water mockup of the THTF heater rod bundle. The studies to date are described

  7. Transient analysis of multifailure conditions by using PWR plant simulator

    International Nuclear Information System (INIS)

    Morisaki, Hidetoshi; Yokobayashi, Masao.

    1984-11-01

    This report describes results of the analysis of abnormal transients caused by multifailures using a PWR plant simulator. The simulator is based on an existing 822MWe power plant with 3 loops, and designed to cover wide range of plant operation from cold shutdown to full power at the end of life. Various malfunctions to simulate abnormal conditions caused by equipment failures are provided. In this report, features of abnormal transients caused by concurrence of malfunctions are discussed. The abnormal conditions studied are leak of primary coolant, loss of charging and feedwater flows, and control systems failure. From the results, it was observed that transient responses caused by some of the malfunctions are almost same as the addition of behaviors caused by each single malfunction. Therefore, it can be said that kinds of malfunctions which are concurrent may be estimated from transient characteristics of each single malfunction. (author)

  8. Behaviour of organic iodides under pwr accident conditions

    International Nuclear Information System (INIS)

    Alm, M.

    1982-01-01

    Laboratory experiments were performed to study the behaviour of radioactive methyl iodide under PWR loss-of-coolant conditions. The pressure relief equipment consisted of an autoclave for simulating the primary circuit and of an expansion vessel for simulating the conditions after a rupture in the reactor coolant system. After pressure relief, the composition of the CH 3 sup(127/131)I-containing steam-air mixture within the expansion vessel was analysed at 80 0 C over a period of 42 days. On the basis of the values measured and of data taken from the literature, both qualitative and quantitative assessments have been made as to the behaviour of radioactive methyl iodide in the event of loss-of-coolant accidents. (author)

  9. The verification of PWR-fuel code for PWR in-core fuel management

    International Nuclear Information System (INIS)

    Surian Pinem; Tagor M Sembiring; Tukiran

    2015-01-01

    In-core fuel management for PWR is not easy because of the number of fuel assemblies in the core as much as 192 assemblies so many possibilities for placement of the fuel in the core. Configuration of fuel assemblies in the core must be precise and accurate so that the reactor operates safely and economically. It is necessary for verification of PWR-FUEL code that will be used in-core fuel management for PWR. PWR-FUEL code based on neutron transport theory and solved with the approach of multi-dimensional nodal diffusion method many groups and diffusion finite difference method (FDM). The goal is to check whether the program works fine, especially for the design and in-core fuel management for PWR. Verification is done with equilibrium core search model at three conditions that boron free, 1000 ppm boron concentration and critical boron concentration. The result of the average burn up fuel assemblies distribution and power distribution at BOC and EOC showed a consistent trend where the fuel with high power at BOC will produce a high burn up in the EOC. On the core without boron is obtained a high multiplication factor because absence of boron in the core and the effect of fission products on the core around 3.8 %. Reactivity effect at 1000 ppm boron solution of BOC and EOC is 6.44 % and 1.703 % respectively. Distribution neutron flux and power density using NODAL and FDM methods have the same result. The results show that the verification PWR-FUEL code work properly, especially for core design and in-core fuel management for PWR. (author)

  10. Plutonium-239 production rate study using a typical fusion reactor

    International Nuclear Information System (INIS)

    Faghihi, F.; Havasi, H.; Amin-Mozafari, M.

    2008-01-01

    The purpose of the present paper is to compute fissile 239 Pu material by supposed typical fusion reactor operation to make the fuel requirement for other purposes (e.g. MOX fissile fuel, etc.). It is assumed that there is a fusion reactor has a cylindrical geometry and uses uniformly distributed deuterium-tritium as fuel so that neutron wall load is taken at 10(MW)/(m 2 ) . Moreover, the reactor core is surrounded by six suggested blankets to make best performance of the physical conditions described herein. We determined neutron flux in each considered blanket as well as tritium self-sufficiency using two groups neutron energy and then computation is followed by the MCNP-4C code. Finally, material depletion according to a set of dynamical coupled differential equations is solved to estimate 239 Pu production rate. Produced 239 Pu is compared with two typical fission reactors to find performance of plutonium breeding ratio in the case of the fusion reactor. We found that 0.92% of initial U is converted into fissile Pu by our suggested fusion reactor with thermal power of 3000 MW. For comparison, 239 Pu yield of suggested large scale PWR is about 0.65% and for LMFBR is close to 1.7%. The results show that the fusion reactor has an acceptable efficiency for Pu production compared with a large scale PWR fission reactor type

  11. Plutonium-239 production rate study using a typical fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Faghihi, F. [Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Havasi, H.; Amin-Mozafari, M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51154 Shiraz (Iran, Islamic Republic of)

    2008-05-15

    The purpose of the present paper is to compute fissile {sup 239}Pu material by supposed typical fusion reactor operation to make the fuel requirement for other purposes (e.g. MOX fissile fuel, etc.). It is assumed that there is a fusion reactor has a cylindrical geometry and uses uniformly distributed deuterium-tritium as fuel so that neutron wall load is taken at 10(MW)/(m{sup 2}) . Moreover, the reactor core is surrounded by six suggested blankets to make best performance of the physical conditions described herein. We determined neutron flux in each considered blanket as well as tritium self-sufficiency using two groups neutron energy and then computation is followed by the MCNP-4C code. Finally, material depletion according to a set of dynamical coupled differential equations is solved to estimate {sup 239}Pu production rate. Produced {sup 239}Pu is compared with two typical fission reactors to find performance of plutonium breeding ratio in the case of the fusion reactor. We found that 0.92% of initial U is converted into fissile Pu by our suggested fusion reactor with thermal power of 3000 MW. For comparison, {sup 239}Pu yield of suggested large scale PWR is about 0.65% and for LMFBR is close to 1.7%. The results show that the fusion reactor has an acceptable efficiency for Pu production compared with a large scale PWR fission reactor type.

  12. Fluid-structure coupled dynamic response of PWR core barrel during LOCA

    International Nuclear Information System (INIS)

    Lu, M.W.; Zhang, Y.G.; Shi, F.

    1991-01-01

    This paper is engaged in the Fluid-Structure Interaction LOCA analysis of the core barrel of PWR. The analysis is performed by a multipurpose computer code SANES. The FSI inside the pressure vessel is treated by a FEM code including some structural and acoustic elements. The transient in the primary loop is solved by a two-phase flow code. Both codes are coupled one another. Some interesting conclusions are drawn. (author)

  13. Safety aspects of the design of a PWR gaseous radwaste treatment system using hydrogen recombiners

    International Nuclear Information System (INIS)

    Glibert, R.; Nuyt, G.; Herin, S.; Fossion, P.

    1978-01-01

    PWR Gaseous radwaste treatment system is essential for the reduction of impact on environment of the nuclear power plants. Decay tank system has been used for the retention of the radioactive gaseous fission products generated in the primary coolant. The use of a system combining decay tanks and hydrogen recombiner units is described in this paper. Accent is put on the safety aspects of this gaseous radwaste treatment facilitystudied by BN for a Belgian Power Plant. (author)

  14. Zinc injection in German PWR plants

    International Nuclear Information System (INIS)

    Streit, K.

    2004-01-01

    Operating experience acquired at PWR NNPs shows that zinc injection at low concentrations of 5 ppb is a very effective source term reduction measure. This method does not lead to any operating restrictions or other negative effects on plant systems and components. The nuclear industry has been very successful in reducing radiation exposures within the past two decades. Annual exposures could be significantly decreased and are now at a level of around 1 man-Sv per plant and year. This great success can mainly be attributed to the general commitment of plant operators to maintaining radiation exposures of workers in the controlled access area as low as reasonably achievable (ALARA principle). The ALARA principle, of course, also implies evaluation of the economic benefit of radiation protection measures. Radiation source term reduction has drawn increasing attention of plant operators in recent years. For the new PWRs cobalt-based alloys in the primary system have successively been eliminated already at the design and construction phase within the last decade. Use of wear-resistant cobalt-free substitute materials in combination with the general use of advanced alloys for the steam generator tubing of PWRs resulted in low values for the two most common sources of plant radiation fields, namely 58 Co and 60 Co. Investigations showed that the beneficial effect of zinc can be related to its high affinity for mixed spinel oxide phases, resulting in the following two basic effects: -Zinc is incorporated preferentially into the oxide layer on primary system surfaces and thus reduces pickup of 58 Co and 60 Co and - Zinc can displace cobalt isotopes from existing oxide layers. In German PWRs with Incoloy 800 steam generator tubing material (Ni-content -32%), the observed reductions correspond to a decrease in dose rates of around 10 to 15% per year and thus follow, as predicted, the half-life time of 60 Co. Overall reductions in high radiation areas are now in the range of

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

    International Nuclear Information System (INIS)

    Algarotti, Marc; Dubois, Philippe; Hernandez, Luc; Landez, Jean Paul

    2006-01-01

    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 requirements of

  16. Simulation of nonlinear dynamics of a PWR core by an improved lumped formulation for fuel heat transfer

    International Nuclear Information System (INIS)

    Su, Jian; Cotta, Renato M.

    2000-01-01

    In this work, thermohydraulic behaviour of PWR, during reactivity insertion and partial loss-of-flow, is simulated by using a simplified mathematical model of reactor core and primary coolant. An improved lumped parameter formulation for transient heat conduction in fuel rod is used for core heat transfer modelling. Transient temperature response of fuel, cladding and coolant is analysed. (author)

  17. PWR reload safety evaluation methodology

    International Nuclear Information System (INIS)

    Doshi, P.K.; Chapin, D.L.; Love, D.S.

    1993-01-01

    The current practice for WWER safety analysis is to prepare the plant Safety Analysis Report (SAR) for initial plant operation. However, the existing safety analysis is typically not evaluated for reload cycles to confirm that all safety limits are met. In addition, there is no systematic reanalysis or reevaluation of the safety analyses after there have been changes made to the plant. The Westinghouse process is discussed which is in contrast to this and in which the SAR conclusions are re-validated through evaluation and/or analysis of each reload cycle. (Z.S.)

  18. PWR blowdown heat transfer separate-effects program: thermal-hydraulic test facility experimental data report for test 104

    International Nuclear Information System (INIS)

    Leon, D.M.; White, M.D.; Moore, P.A.; Hedrick, R.A.

    1978-01-01

    Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 104, which is part of the ORNL Pressurized-Water Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant accident in the PWR system. Test 104 was conducted to obtain CHF in bundle 1 under blowdown conditions. The primary purpose of this report is to make the reduced instrument responses during test 104 available

  19. Development of computational program for studying the reactor control system in PWR plants; Desenvolvimento de um programa computacional para estudo do sistema de controle do reator em plantas PWR

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Jose Ricardo de; Soares, Adalberto Jose [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)

    2002-07-01

    In this work a computational program is presented which has been developed for specific application on the study of the reactor control system of a typical PWR plant. As to the basic function of simulating power transients the program has the following structure: a representative mathematical model of the dynamic and stationary behaviors of the primary circuit; a group of equations associated to the reactor power control and system pressure control; screens for the entry of reference data as well as of control blocks and control bar speed programming module parameters; main entering screens for the configuration of the excitement/transient function as well as of simulation time and control mood; and graphical output of all the process variables incorporated to the model. As premise it has been considered as sufficient the modeling of the primary circuit, a differential equation being used which associates the average temperature of the coolant within the steam generator with the potency transferred to the secondary circuit, denominated 'secondary potency', as an interface with the secondary circuit. Every transient - ramp or step - is established upon the 'turbine power' variable, which in turn is related to the 'secondary power' variable by means of a differential equation that represents a first - order delay, having adjustable parameters on the data - entry screen. In the neutronic model as defined for the reactor, the reactivity feedback effects due to primary circuit pressure variation, as well as fuel and coolant temperature variation, were taken into consideration. Thermo-hydraulics constants and project data taken from the available bibliography, adapted to a particular small PWR unit conception , were employed for loading the program. With the open-loop simulation results a positive qualitative evaluation of the program was obtained, in comparison to published results related to simulators bearing equal purposes, more

  20. Study of PWR reactor efficiency as a function of turbine steam extractions

    International Nuclear Information System (INIS)

    Rocha, Janine Gandolpho da; Alvim, Antonio Carlos Marques; Martinez, Aquilino Senra

    2002-01-01

    The objective of this work is to optimize the extractions of the low-pressure turbine of a PWR nuclear reactor, in order to obtain the best thermodynamic cycle efficiency. We have analyzed typical data of a 1300 MW PWR reactor, operating at 25%, 50%, 75% and 100% capacities, respectively. The first stage of this study consists of generating a mathematical model capable of describing the reactor behavior and efficiency at any power level. The second stage of this study consists of to combine the generated mathematical model in an optimization computer program that optimize the extractions flow of the low-pressure turbine until it finds the optimal system efficiency. This work does not alter the nuclear facility project in any way. (author)

  1. 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.

  2. PWR station blackout transient simulation in the INER integral system test facility

    International Nuclear Information System (INIS)

    Liu, T.J.; Lee, C.H.; Hong, W.T.; Chang, Y.H.

    2004-01-01

    Station blackout transient (or TMLB' scenario) in a pressurized water reactor (PWR) was simulated using the INER Integral System Test Facility (IIST) which is a 1/400 volumetrically-scaled reduce-height and reduce-pressure (RHRP) simulator of a Westinghouse three-loop PWR. Long-term thermal-hydraulic responses including the secondary boil-off and the subsequent primary saturation, pressurization and core uncovery were simulated based on the assumptions of no offsite and onsite power, feedwater and operator actions. The results indicate that two-phase discharge is the major depletion mode since it covers 81.3% of the total amount of the coolant inventory loss. The primary coolant inventory has experienced significant re-distribution during a station blackout transient. The decided parameter to avoid the core overheating is not the total amount of the coolant inventory remained in the primary core cooling system but only the part of coolant left in the pressure vessel. The sequence of significant events during transient for the IIST were also compared with those of the ROSA-IV large-scale test facility (LSTF), which is a 1/48 volumetrically-scaled full-height and full-pressure (FHFP) simulator of a PWR. The comparison indicates that the sequence and timing of these events during TMLB' transient studied in the RHRP IIST facility are generally consistent with those of the FHFP LSTF. (author)

  3. System for stress corrosion conditions tests on PWR reactors

    International Nuclear Information System (INIS)

    Castro, Andre Cesar de Jesus

    2007-01-01

    The study of environmentally assisted cracking (EAC) involves the consideration and evaluation of the inherent compatibility between a material and the environment under conditions of either applied or residual stress. EAC is a critical problem because equipment, components and structure are subject to the influence of mechanical stress, water environment of different composition, temperature and different material history. Testing for resistance to EAC is one of the most effective ways to determine the interrelationships among this variables on the process of EAC. Up to now, several experimental techniques have been developed worldwide, which address different aspects of environmental caused damage. Constant loading of CT specimens test is a typical example of test, which is used for the estimation of parameters of stress corrosion cracking. To assess the initiation stages and kinetics of crack growth, the testing facility should allow active loading of specimens in the environment that is close to the actual operation conditions of assessed component. This paper presents a testing facility for stress corrosion cracking to be installed at CDTN, which was designed and developed at CDTN. The facility is used to carry out constant load tests under simulated PWR environment, where temperature, water pressure and chemistry are controlled, which are considered the most important factors in SCC. Also, the equipment operational conditions, its applications, and restrictions are presented. The system was developed to operate at temperature until 380 degree C and pressure until 180 bar. It consists in a autoclave stuck at a mechanical system, responsible of producing load , a water treatment station, and a data acquisition system. This testing facility allows the evaluation of cracking progress, especially at PWR reactor. (author) operational conditions. (author)

  4. RSK-guidelines for PWR reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The RSK guidelines for PWA reactors of April 24, 1974, have been revised and amended in this edition. The RSK presents a summary of safety requirements to be observed in the design, construction, and operation of PWR reactors in the form of guidelines. From January 1979 onwards these guidelines will be the basis of siting and safety considerations for new PWR reactors, and newly built nuclear power plants will have to form these guidelines. They are not binding for existing nuclear power plants under construction or in operation. It will be a matter of individual discussion whether or not the guidelines will be applied in these plants. The main purpose of the guidelines is to facilitate discussion among RSK members and to give early information on necessary safety requirements. If the guidelines are observed by producers and operators, the RSK will make statements on individual projects at short notice. (orig./HP) [de

  5. Optimization of reload core design for PWR

    International Nuclear Information System (INIS)

    Shen Wei; Xie Zhongsheng; Yin Banghua

    1995-01-01

    A direct efficient optimization technique has been effected for automatically optimizing the reload of PWR. The objective functions include: maximization of end-of-cycle (EOC) reactivity and maximization of average discharge burnup. The fuel loading optimization and burnable poison (BP) optimization are separated into two stages by using Haling principle. In the first stage, the optimum fuel reloading pattern without BP is determined by the linear programming method using enrichments as control variable, while in the second stage the optimum BP allocation is determined by the flexible tolerance method using the number of BP rods as control variable. A practical and efficient PWR reloading optimization program based on above theory has been encoded and successfully applied to Qinshan Nuclear Power Plant (QNP) cycle 2 reloading design

  6. PWR fuel behavior: lessons learned from LOFT

    International Nuclear Information System (INIS)

    Russell, M.L.

    1981-01-01

    A summary of the experience with the Loss-of-Fluid Test (LOFT) fuel during loss-of-coolant experiments (LOCEs), operational and overpower transient tests and steady-state operation is presented. LOFT provides unique capabilities for obtaining pressurized water reactor (PWR) fuel behavior information because it features the representative thermal-hydraulic conditions which control fuel behavior during transient conditions and an elaborate measurement system to record the history of the fuel behavior

  7. EDF/CIDEN - ONECTRA: PWR decontamination

    International Nuclear Information System (INIS)

    Fayolle, P.; Orcel, H.; Wertz, L.

    2010-01-01

    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

  8. Optimum fuel use in PWR reactors

    International Nuclear Information System (INIS)

    Neubauer, W.

    1979-07-01

    An optimization program was developed to calculate minimum-cost refuelling schedules for PWR reactors. Optimization was made over several cycles, without any constraints (equilibrium cycle). In developing the optimization program, special consideration was given to an individual treatment of every fuel element and to a sufficiently accurate calculation of all the data required for safe reactor operation. The results of the optimization program were compared with experimental values obtained at Obrigheim nuclear power plant. (orig.) [de

  9. Chemical and radiochemical specifications - PWR power plants

    International Nuclear Information System (INIS)

    Stutzmann, A.

    1997-01-01

    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. GAIA: AREVAs New PWR fuel assembly design

    Energy Technology Data Exchange (ETDEWEB)

    Vollmert, N.; Gentet, G.; Louf, P.H.; Mindt, M.; O' Brian, J.; Peucker, J.

    2015-07-01

    GAIA is the label of a new PWR Fuel Assembly design developed by AREVA with the objective to provide its customers an advanced fuel assembly design regarding both robustness and performance. Since 2012 GAIA lead fuel assemblies are under irradiation in a Swedish reactor and since 2015 in a U.S. reactor. Visual inspections and examinations carried out so far during the outages confirmed the intended reliability, robustness and the performance enhancement of the design. (Author)

  11. Shielding design for PWR in France

    International Nuclear Information System (INIS)

    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

  12. Sensitivity analysis of a PWR pressurizer

    International Nuclear Information System (INIS)

    Bruel, Renata Nunes

    1997-01-01

    A sensitivity analysis relative to the parameters and modelling of the physical process in a PWR pressurizer has been performed. The sensitivity analysis was developed by implementing the key parameters and theoretical model lings which generated a comprehensive matrix of influences of each changes analysed. The major influences that have been observed were the flashing phenomenon and the steam condensation on the spray drops. The present analysis is also applicable to the several theoretical and experimental areas. (author)

  13. T Plant removal of PWR Chiller Subsystem

    International Nuclear Information System (INIS)

    Dana, C.M.

    1994-01-01

    The PWR Pool Chiller System is not longer required for support of the Shippingport Blanket Fuel Assemblies Storage. The Engineering Work Plan will provide the overall coordination of the documentation and physical changes to deactivate the unneeded subsystem. The physical removal of all energy sources for the Chiller equipment will be covered under a one time work plan. The documentation changes will be covered using approved Engineering Change Notices and Procedure Change Authorizations as needed

  14. Nondestructive examination requirements for PWR vessel internals

    International Nuclear Information System (INIS)

    Spanner, J.

    2015-01-01

    This paper describes the requirements for the nondestructive examination of pressurized water reactor (PWR) vessel internals in accordance with the requirements of the EPRI Material Reliability Program (MRP) inspection standard for PWR internals (MRP-228) and the American Society of Mechanical Engineers Section XI In-service Inspection. The MRP vessel internals examinations have been performed at nuclear plants in the USA since 2009. The objective of the inspection standard is to provide the requirements for the nondestructive examination (NDE) methods implemented to support the inspection and evaluation of the internals. The inspection standard contains requirements specific to the inspection methodologies involved as well as requirements for qualification of the NDE procedures, equipment and personnel used to perform the vessel internals inspections. The qualification requirements for the NDE systems will be summarized. Six PWR plants in the USA have completed inspections of their internals using the Inspection and Evaluation Guideline (MRP-227) and the Inspection Standard (MRP-228). Examination results show few instances of service-induced degradation flaws, as expected. The few instances of degradation have mostly occurred in bolting

  15. The Conceptual Design of Innovative Safe PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Han-Gon [Centural Research Institute, Daejeon (Korea, Republic of); Heo, Sun [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2016-10-15

    Most of countries operating NPPs have been performed post-Fukushima improvements as short-term countermeasure to enhance the safety of operating NPPs. Separately, vendors have made efforts on developing passive safety systems as long-term and ultimate countermeasures. AP1000 designed by Westinghouse Electric Company has passive safety systems including the passive emergency core cooling system (PECCS), the passive residual heat removal system (PRHRS), and the passive containment cooling system (PCCS). ESBWR designed by GE-Hitachi also has passive safety systems consisting of the isolation condenser system, the gravity driven cooling system and the PCCS. Other countries including China and Russia have made efforts on developing passive safety systems for enhancing the safety of their plants. In this paper, we summarize the design goals and main design feature of innovative safe PWR, iPOWER which is standing for Innovative Passive Optimized World-wide Economical Reactor, and show the developing status and results of research projects. To mitigate an accident without electric power and enhance the safety level of PWR, the conceptual designs of passive safety system and innovative safe PWR have been performed. It includes the PECCS for core cooling and the PCCS for containment cooling. Now we are performing the small scale and separate effect tests for the PECCS and the PCCS and preparing the integral effect test for the PECCS and real scale test for the PCCS.

  16. PWR water chemistry controls: a perspective on industry initiatives and trends relative to operating experience and the EPRI PWR water chemistry guidelines

    International Nuclear Information System (INIS)

    Fruzzetti, K.; Choi, S.; Haas, C.; Pender, M.; Perkins, D.

    2010-01-01

    An effective PWR water chemistry control program must address the following goals: Minimize materials degradation (e.g., PWSCC, corrosion of fuel, corrosion damage of steam generator (SG) tubes); Maintain fuel integrity and good performance; Minimize corrosion product transport (e.g., transport and deposition on the fuel, transport into the SGs where it can foul tube surfaces and create crevice environments for the concentration of corrosive impurities); Minimize dose rates. Water chemistry control must be optimized to provide overall improvement considering the sometimes variant constraints of the goals listed above. New technologies are developed for continued mitigation of materials degradation, continued fuel integrity and good performance, continued reduction of corrosion product transport, and continued minimization of plant dose rates. The EPRI chemistry program, in coordination with other EPRI programs, strives to improve these areas through application of chemistry initiatives, focusing on these goals. This paper highlights the major initiatives and issues with respect to PWR primary and secondary system chemistry and outlines the recent, on-going, and proposed work to effectively address them. These initiatives are presented in light of recent operating experience, as derived from EPRI's PWR chemistry monitoring and assessment program, and EPRI's water chemistry guidelines. (author)

  17. PWR.2 - the unique transportation

    International Nuclear Information System (INIS)

    Howell, G.E.; Wills, L.I.

    1987-01-01

    Design studies of the prototype machinery and installation of same to be used for test and evaluation of a new design of nuclear power plant for submarines, showed that there were advantages if large units could be fitted out entirely at the manufacturer's base in Barrow-in-Furness. However, they had then to be shipped to the customer at the Vulcan Naval Reactor Test Establishment in Caithness, Scotland. The transportation of the loads involved is described. The main loads were the primary unit which weighed 1300 tonnes and the secondary unit which was transported as five separate assemblies, the largest two of which weighed 151 and 43 tonnes. Five basic transportation methods were used: skidding on PTFE pads, sea passage with barge on submersible barge, rolling on airbags, skating on water skates and lifting and rolling on multi-wheeled trailers. By careful planning the primary unit was moved in 16 days and the secondary unit in 19 days. The route and methods used are described and illustrated. (U.K.)

  18. Program of monitoring PWR fuel in Spain; Programa de Vigilancia de Combustible pwr en Espana

    Energy Technology Data Exchange (ETDEWEB)

    Martinez Murillo, J. C.; Quecedo, M.; Munoz-Roja, C.

    2015-07-01

    In the year 2000 the PWR utilities: Centrales Nucleares Almaraz-Trillo (CNAT) and Asociacion Nuclear Asco-Vandellos (ANAV), and ENUSA Industrias Avanzadas developed and executed a coordinated strategy named PIC (standing for Coordinated Research Program), for achieving the highest level of fuel reliability. The paper will present the scope and results of this program along the years and will summarize the way the changes are managed to ensure fuel integrity. The excellent performance of the ENUSA manufactured fuel in the PWR Spanish NPPs is the best indicator that the expectations on this program are being met. (Author)

  19. Thermo-mechanical analysis of PWR bolts susceptible to IASCC

    International Nuclear Information System (INIS)

    Matteoli, C.; Hannink, M.H.C.; Blom, F.J.; Marck, S.C. van der; Charpin-Jacobs, F.

    2015-01-01

    Irradiation Assisted Stress Corrosion Cracking (IASCC) is considered a primary ageing issue for the Reactor Pressure Vessel (RPV) internals of Pressurized Water Reactors (PWR). In particular, this complex phenomenon which develops in an environment featuring thermal and mechanical stresses, interaction with corrosive compounds and irradiation, is affecting the bolts connecting the baffles and the formers in the Nuclear Power Plants' RPVs. The baffle-former assembly is the structure that borders the fuel assemblies region, contributing to keep them in position and separating in the radial direction, the core region from the downcomer region. An evaluation of the stresses and temperatures reached in the baffle-former bolts during normal operation was performed by means of a coupled thermo-mechanical study which uses reactor physics calculations to obtain the fluence in the reactor core and as a consequence the heat deposition in the RPV internals. The heat deposition data are coupled with a finite element model of the bolts and the RPV internals in order to perform a complete analysis taking in account thermal, mechanical and radiation loadings. The study is first carried out focusing on a section of the RPV internals, showing a single row of baffle-former bolts. Then the work is extended to the full core height. The model set up in this work, includes an in-depth study of the behavior of the core internals, in particular baffle-former bolts. The model has the capability of understanding the mechanical and thermal behavior of essential internal components in a PWR. (authors)

  20. PWR and BWR spent fuel assembly gamma spectra measurements

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-11

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

  1. Zircaloy-4 corrosion in PWR's

    International Nuclear Information System (INIS)

    Fyfitch, S.; Smalley, W.R.; Roberts, E.

    1985-01-01

    Zircaloy-4 waterside corrosion has been studied extensively in the nuclear industry for a number of years. Following the early crud-related corrosion failures in the Saxton test reactor, Westinghouse undertook numerous programs to minimize crud deposition on fuel rods in power reactors through primary coolant chemistry control. Modern plants today are operating with improved coolant chemistry guidelines, and crud deposition levels are very low in proportion to earlier experience. Zircaloy-4 corrosion under a variety of coolant chemistry, heat flux and exposure conditions has been studied extensively. Experience to date, even in relatively high coolant temperature plants, has indicated that -for both fuel cladding and structural components- Zircaloy-4 waterside corrosion performance has been excellent. Recognizing future industry trends, however, which will result in Zircaloy-4 being subjected to ever increasing corrosion duties, Westinghouse will continue accumulating Zircaloy-4 corrosion experience in large power plants. 13 refs.

  2. PWR type process heat reactor

    International Nuclear Information System (INIS)

    Aubert, Gilles; Petit, Guy.

    1974-01-01

    The nuclear reactor described is of the pressurized water type. It includes a prestressed concrete vessel, the upper part of which is shut by a closure, and a core surrounded by a core ring. The core fuel assemblies are supported by an initial set of vertical tubes integral with the bottom of the vessel, which serve to guide the rods of the control system. Over the core there is a second set of vertical tubes, able to receive the absorbing part of a control rod when this is raised above the core. An annular pressurizer around the core ring keeps the water in a liquid state. A pump is located above the second set of tubes and is integral with the closure. It circulates the water between the core and the intake of at least one primary heat exchanger, the exchanger (s) being placed between the wall of the vessel and the core ring [fr

  3. Water chemistry control of PWR nuclear power plant

    International Nuclear Information System (INIS)

    Hino, Yuichi; Makino, Ichiro; Yamauchi, Sumio; Fukuda, Fumihito.

    1992-01-01

    In PWR power plants, the primary system taking heat out of nuclear reactors and the secondary system generating steam and driving turbines are completely separated by steam generators, accordingly, by mutually independent water treatment, both systems are to be maintained in the optimal conditions. Namely, primary system is the closed water circulation circuit of simple liquid phase though under high temperature, high pressure condition, therefore, water shows the stable physical and chemical properties, and the minute water treatment for restraining the corrosion of structural materials and reducing radioactivity can be done. Secondary system is similar to the condensate and feedwater system of thermal power plants, and is the circuit for liquid-vapor two-phase transformation, but due to the local concentration of impurities by evaporation, the strict requirement is set for secondary water quality. However, secondary system can be treated in the state without radioactivity, and this is a great merit. The outline, basic concept and execution of primary water quality control, and the outline, concept, control criteria, facilities and execution of secondary water quality control are reported. (K.I.)

  4. Control in fabrication of PWR and BWR type reactor fuel elements

    International Nuclear Information System (INIS)

    Gorskij, V.V.

    1981-01-01

    Both destructive and non-destructive testing methods now in use in fabrication of BWR and PWR type reactor fuel elements at foreign plants are reviewed. Technological procedures applied in fabrication of fuel elements and fuel assemblies are described. Major attention is paid to radiographic, ultrasonic, metallographic, visual and autoclavic testings. A correspondence of the methods applied to the ASTM standards is discussed. The most part of the countries are concluded the apply similar testing methods enabling one to reliably evaluate the quality of primary materials and fabricated fuel elements and thus meeting the demands to contemporary PWR and BWR type reactor fuel elements. Practically all fuel element and pipe fabrication plants in Western Europe, Asia and America use the ASTM standards as the basis for the quality contr [ru

  5. Preliminary assessment of PWR Steam Generator modelling in RELAP5/MOD3

    International Nuclear Information System (INIS)

    Preece, R.J.; Putney, J.M.

    1993-07-01

    A preliminary assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD3 is presented. The study is based on calculations against a series of steady-state commissioning tests carried out on the Wolf Creek PWR over a range of load conditions. Data from the tests are used to assess the modelling of primary to secondary side heat transfer and, in particular, to examine the effect of reverting to the standard form of the Chen heat transfer correlation in place of the modified form applied in RELAP5/MOD2. Comparisons between the two versions of the code are also used to show how the new interphase drag model in RELAP5/MOD3 affects the calculation of SG liquid inventory and the void fraction profile in the riser

  6. Pu recycling in a full Th-MOX PWR core. Part I: Steady state analysis

    International Nuclear Information System (INIS)

    Fridman, E.; Kliem, S.

    2011-01-01

    Research highlights: → Detailed 3D 100% Th-MOX PWR core design is developed. → Pu incineration increased by a factor of 2 as compared to a full MOX PWR core. → The core controllability under steady state conditions is demonstrated. - Abstract: Current practice of Pu recycling in existing Light Water Reactors (LWRs) in the form of U-Pu mixed oxide fuel (MOX) is not efficient due to continuous Pu production from U-238. The use of Th-Pu mixed oxide (TOX) fuel will considerably improve Pu consumption rates because virtually no new Pu is generated from thorium. In this study, the feasibility of Pu recycling in a typical pressurized water reactor (PWR) fully loaded with TOX fuel is investigated. Detailed 3-dimensional 100% TOX and 100% MOX PWR core designs are developed. The full MOX core is considered for comparison purposes. The design stages included determination of Pu loading required to achieve 18-month fuel cycle assuming three-batch fuel management scheme, selection of poison materials, development of the core loading pattern, optimization of burnable poison loadings, evaluation of critical boron concentration requirements, estimation of reactivity coefficients, core kinetic parameters, and shutdown margin. The performance of the MOX and TOX cores under steady-state condition and during selected reactivity initiated accidents (RIAs) is compared with that of the actual uranium oxide (UOX) PWR core. Part I of this paper describes the full TOX and MOX PWR core designs and reports the results of steady state analysis. The TOX core requires a slightly higher initial Pu loading than the MOX core to achieve the target fuel cycle length. However, the TOX core exhibits superior Pu incineration capabilities. The significantly degraded worth of control materials in Pu cores is partially addressed by the use of enriched soluble boron and B 4 C as a control rod absorbing material. Wet annular burnable absorber (WABA) rods are used to flatten radial power distribution

  7. Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

    International Nuclear Information System (INIS)

    Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C.; Palma, Daniel A.P.

    2017-01-01

    A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

  8. Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C., E-mail: sabrinapral@gmail.com, E-mail: amir@cdtn.brm, E-mail: hcr@cdtn.br, E-mail: hcr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Junjie [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); Lu, Zhanpeng, E-mail: zplu@t.shu.edu.cn [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); State Key Laboratory of Advanced Special Steels, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Xiao, Qian; Ru, Xiangkun; Han, Guangdong; Chen, Zhen [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); Zhou, Bangxin [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); State Key Laboratory of Advanced Special Steels, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Shoji, Tetsuo [New Industry Creation Hatchery Center, Tohoku University, Sendai 980-8579 (Japan)

    2016-04-15

    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. - Highlights: • Transgranular fatigue crack growth rate was not affected by the cold rolling orientation. • Locally intergranular SCC was found in the hydrogenated PWR water. • Extensive intergranular SCC cracks were found in deaerated PWR water. • T–L specimen showed more extensive SCC cracks and a higher crack growth rate. • Crack branching related to the applied stress and the preferential oxidation path.

  10. Study of anticipated transient without scram for PWR

    International Nuclear Information System (INIS)

    Pu Jilong.

    1985-01-01

    Anticipated Transient Without Scram (ATWS) of PWR, the one of the 'Unresolved Safety Issue' with NRC, has been investigated for many years. The latest analysis done by the author considers the PWR's inherent stability and long-term performence under the condition of ATWS combined with SBLOCA and studies the sensitivity of several assumptions, which shows positive results

  11. Pushing back the boundaries of PWR fuel performance

    International Nuclear Information System (INIS)

    Sofer, G.A.; Skogen, F.B.; Brown, C.A.; Fresk, Y.U.

    1985-01-01

    In today's fiercely competitive PWR reload market utilities are benefiting from a variety of design innovations which are helping to cut fuel cycle costs and to improve fuel performance. An advanced PWR fuel design from Exxon, for example, currently under evaluation at the Ginna plant in the United States, offers higher burn-up and greater power cycling. (author)

  12. Highlights of the French program on PWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Pages, J P [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    1997-12-01

    The presentation reviews the French programme on PWR fuel including the overall results of the year 1996 for nuclear operation; fuel management and economy; French nuclear electricity generation sites; production of nuclear generated electricity; energy availability of the 900 and 1,300 Mw PWR units; average radioactive liquid releases excluding tritium per unit; plutonium recycling experience.

  13. An economic analysis code used for PWR fuel cycle

    International Nuclear Information System (INIS)

    Liu Dingqin

    1989-01-01

    An economic analysis code used for PWR fuel cycle is developed. This economic code includes 12 subroutines representing vavious processes for entire PWR fuel cycle, and indicates the influence of the fuel cost on the cost of the electricity generation and the influence of individual process on the sensitivity of the fuel cycle cost

  14. Sizewell: proposed site for Britain's first PWR power station

    International Nuclear Information System (INIS)

    1980-10-01

    The pamphlet covers the following points, very briefly: nuclear power - a success story; the Government's nuclear programme; why Sizewell; the PWR (with diagram); the PWR at Sizewell (with aerial view) (location; size; cooling water; road access; fuel transport; construction; employment; environment; screening; the next steps (licensing procedures, etc.); safety; further information). (U.K.)

  15. Chemical decontamination solutions: Effects on PWR equipment

    International Nuclear Information System (INIS)

    Pezze, C.M.; Colvin, E.R.; Aspden, R.G.

    1992-01-01

    A critical objective for the nuclear industry is the reduction of personnel exposure to radiation. Reductions have been achieved through industry's radiation management programs including training and radiation awareness concepts. Increased plant maintenance and higher radiation fields at many sites continue to raise concerns. To alleviate the radiation exposure problem, the sources of radiation which contribute to personnel exposure must be removed from the plant. A feasible was of significantly reducing these sources from a Pressurized Water Reactor (PWR) is to chemically decontaminate the entire reactor coolant system (RCS). A program was conducted to determine the technical acceptability of using certain dilute chemical solvent processes for full RCS chemical decontamination. The two processes evaluated were CAN-DEREM and LOMI. The purpose of the program was to define and complete a systematic evaluation of the major issues that need to be addressed for the successful decontamination of the entire RCS and affected portions of the auxiliary systems of a four-loop PWR system. A test program was designed to evaluate the corrosion effects of the two decontamination processes under expected plant conditions. Materials and sample configurations dictated by generic PWR components were evaluated. The testing also included many standard corrosion coupons. The test data were then used to assess the impact of chemical decontamination on the physical condition and operability of the components, equipment and mechanical systems that make up the RCS. An overview of the test program, sample configurations, data and engineering evaluations is presented. The data demonstrate that through detailed engineering evaluations of corrosion data and equipment function, the impact of full RCS chemical decontamination on plant equipment is established

  16. Transient study of a PWR pressurizer

    International Nuclear Information System (INIS)

    Sotoma, H.

    1973-01-01

    An appropriate method for the calculation and transient performance of the pressurizer of a pressurized water reactor is presented. The study shows a digital program of simulation of pressurizer dynamics based on the First Law of Thermodynamic and Laws of Heat and Mass Transfer. The importance of the digital program that was written for a pressurizer of PWR, lies in the fact that, this can be of practical use in the safety analysis of a reactor of Angra dos Reis type with a power of about 500 M We. (author)

  17. Technical specifications for PWR secondary water chemistry

    International Nuclear Information System (INIS)

    Weeks, J.R.; van Rooyen, D.

    1977-08-01

    The bases for establishing Technical Specifications for PWR secondary water chemistry are reviewed. Whereas extremely stringent control of secondary water needs to be maintained to prevent denting in some units, sound bases for establishing limits that will prevent stress corrosion, wastage, and denting do not exist at the present time. This area is being examined very thoroughly by industry-sponsored research programs. Based on the evidence available to date, short term control limits are suggested; establishment of these or other limits as Technical Specifications is not recommended until the results of the research programs have been obtained and evaluated

  18. Technical basis for PWR emergency plans forming

    International Nuclear Information System (INIS)

    L'Homme, A.; Manesse, D.; Gauvain, J.; Crabol, B.

    1989-10-01

    Our speech first summarizes the french approach concerning the management of severe accidents which could occur on PWR stations. Then it defines the source-term which is being used as a general support for elaborating the emergency plans devoted to the protection of the population. It describes next the consequences of this source-term on the site and in the environment, which constitute the technical bases for defining actions of utilities and concerned authorities. It gives lastly information on the present status of the different emergency plans and the complementary work undertaken to improve them [fr

  19. Industrywide survey of PWR organics. Final report

    International Nuclear Information System (INIS)

    Richards, J.E.; Byers, W.A.

    1986-07-01

    Thirteen Pressurized Water reactor (PWR) secondary cycles were sampled for organic acids, total organic carbon, and inorganic anions. The distribution and removal of organics in a makeup water treatment system were investigted at an additional plant. TOC analyses were used for the analysis of makeup water systems; anion ion chromatography and ion exclusion chromatography were used for the analysis of secondary water systems. Additional information on plant operation and water chemistry was collected in a survey. The analytical and survey data were compared and correlations made

  20. Microcomputer simulation of PWR power plant pressurizer

    International Nuclear Information System (INIS)

    Araujo, L.R.A. de; Calixto Neto, J.; Martinez, A.S.; Schirru, R.

    1990-01-01

    It is presented a method for the simulation of the pressurizer behavior of a PWR power plant. The method was implanted in a microcomputer, and it considers all the devices for the pressure control (spray and relief valves, heaters, controller, etc.). The physical phenomena and the PID (Proportional + Integral + Derivative) controller were mathematically represented by linear relations, uncoupled, discretized in the time. There are three different algorithms which take into account the non-linear effects introduced by the variation of the physical properties due to the temperature and pressure, and also the mutual effects between the physical phenomena and the PID controller. (author)

  1. Minimization of PWR reactor control rods wear

    International Nuclear Information System (INIS)

    Ponzoni Filho, Pedro; Moura Angelkorte, Gunther de

    1995-01-01

    The Rod Cluster Control Assemblies (RCCA's) of Pressurized Water Reactors (PWR's) have experienced a continuously wall cladding wear when Reactor Coolant Pumps (RCP's) are running. Fretting wear is a result of vibrational contact between RCCA rodlets and the guide cards which provide lateral support for the rodlets when RCCA's are withdrawn from the core. A procedure is developed to minimize the rodlets wear, by the shuffling and axial reposition of RCCA's every operating cycle. These shuffling and repositions are based on measurement of the rodlet cladding thickness of all RCCA's. (author). 3 refs, 2 figs, 2 tabs

  2. PWR life time: the EDF project

    International Nuclear Information System (INIS)

    Noel, R.; Reynes, L.; Mercier, J.P.

    1987-01-01

    Operating a very large number of standardized PWR units which supply today 70% of French power generation, Electricite de France is highly interested in getting the best estimate of the safe and economical life of these plants. An extensive program of work has been undertaken in this respect. The studies have first to go through all available data on aging process, survey and maintenance of a limited number of major components. This review will lead to recommendation of complementary work in these fields. The first conclusions are that these units are able to perform a long service time, under provision of careful survey and maintenance [fr

  3. Environmental surveillance of PWR power stations

    International Nuclear Information System (INIS)

    Conti, M.

    1980-01-01

    The action of Electricite de France with respect to the environment of PWR nuclear power stations is essentially centred on prevention. Controls are carried out at two levels: - before the power station goes on stream (radioecological study), - when the power station is operational. The purpose of the controls effected on the radioactive effluents and the environment is to check that the maximum discharge rate stipulated in the corresponding orders is complied with and to ensure that there are no anomalies in the environment [fr

  4. Advancing PWR fuel to meet customer needs

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, F W

    1987-03-01

    Since the introduction of the Optimized Fuel Assembly (OFA) for PWRs in the late 1970s, Westinghouse has continued to work with the utility customers to identify the greatest needs for further advance in fuel performance and reliability. The major customer requirements include longer fuel cycle at lower costs, increased fuel discharge burn-up, enhanced operating flexibility, all accompanied by even greater reliability. In response to these needs, Westinghouse developed Vantage 5 PWR fuel. To optimize reactor operations, Vantage 5 fuel features distinct advantages: integral fuel burnable absorbers, axial and radial blankets, intermediate flow mixers, a removable top nozzle, and assembly modifications to accommodate increased discharge burn-up.

  5. An immersed body method for coupled neutron transport and thermal hydraulic simulations of PWR assemblies

    International Nuclear Information System (INIS)

    Jewer, S.; Buchan, A.G.; Pain, C.C.; Cacuci, D.G.

    2014-01-01

    Highlights: • A new method of coupled radiation transport, heat and momentum exchanges on fluids, and heat transfer simulations. • Simulation of the thermal hydraulics and radiative properties within whole PWR assemblies. • An immersed body method for modelling complex solid domains on practical computational meshes. - Abstract: A recently developed immersed body method is adapted and used to model a typical pressurised water reactor (PWR) fuel assembly. The approach is implemented with the numerical framework of the finite element, transient criticality code, FETCH which is composed of the neutron transport code, EVENT, and the CFD code, FLUIDITY. Within this framework the neutron transport equation, Navier–Stokes equations and a fluid energy conservation equation are solved in a coupled manner on a coincident structured or unstructured mesh. The immersed body method has been used to model the solid fuel pins. The key feature of this method is that the fluid/neutronic domain and the solid domain are represented by overlapping and non-conforming meshes. The main difficulty of this approach, for which a solution is proposed in this work, is the conservative mapping of the energy and momentum exchange between the fluid/neutronic mesh and the solid fuel pin mesh. Three numerical examples are presented which include a validation of the fuel pin submodel against an analytical solution; an uncoupled (no neutron transport solution) PWR fuel assembly model with a specified power distribution which was validated against the COBRA-EN subchannel analysis code; and finally a coupled model of a PWR fuel assembly with reflective neutron boundary conditions. Coupling between the fluid and neutron transport solutions is through the nuclear cross sections dependence on Doppler fuel temperature, coolant density and temperature, which was taken into account by using pre-calculated cross-section lookup tables generated using WIMS9a. The method was found to show good agreement

  6. Fatigue analysis of a PWR steam generator tube sheet

    International Nuclear Information System (INIS)

    Billon, F.; Buchalet, C.; Poudroux, G.

    1985-01-01

    The fatigue analysis of a PWR steam generator (S.G) tube sheet is threefold. First, the flow, pressure and temperature variations during the design transients are defined for both the primary fluid and the normal and auxiliary feedwater. Second, the flow, velocities, pressure and temperature variations of the secondary fluid at the bottom of the downcomer and above the tube sheet are determined for the transients considered. Finally, the corresponding temperatures and stresses in the tube sheet are calculated and the usage factors determined at various locations in the tube sheet. The currently available standard design transients for the primary fluid and the feedwater are too conservative to be utilized as such in the fatigue analysis of the S.G. tube sheets. Thus, a detailed examination and reappraisal of each operating transient was performed. The revised design conditions are used as inputs to the calculation model TEMPTRON. TEMPTRON determines the mixing conditions between the feedwater and the recirculation fluid from the S.G. feedwater nozzles to the center of the tube sheet via the downcomer. The fluid parameters, flow rate and velocity, temperature and pressure variations, as a function of the time during the transients are obtained. Finally, the usage factors at various locations on the tube sheet are derived using the standard ASME section III method

  7. Thermohydraulic status and component behavior in the PWR during the selected meltdown scenario station blackout (SBO); Thermohydraulisches Verhalten und Komponentenverhalten eines DWR bei ausgewaehltem Kernschmelzszenarium infolge Station Blackout (SBO). Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Band, Sebastian; Blaesius, Christoph; Scheuerer, Martina; Steinroetter, Thomas

    2017-09-15

    The report on the thermohydraulic status and component behavior in the PWR during the selected meltdown scenario station blackout (SBO) includes the following issues: status of science and technology on this topic, analysis of a high-pressure meltdown scenario using ATHLET-CD for a German PWR starting from the initiating event station blackout, three-dimensional computational fluid dynamic (CFD) analyses of the pressurizer coolant loop in a generic German PWR, evaluation of the thermohydraulic steam generator behavior and its effect on the involved primary circuit components.

  8. Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes

    International Nuclear Information System (INIS)

    Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

    2016-01-01

    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)

  9. CECP, Decommissioning Costs for PWR and BWR

    International Nuclear Information System (INIS)

    Bierschbach, M.C.

    1997-01-01

    1 - Description of program or function: The Cost Estimating Computer Program CECP, designed for use on an IBM personal computer or equivalent, was developed for estimating the cost of decommissioning boiling water reactor (BWR) and light-water reactor (PWR) power stations to the point of license termination. 2 - Method of solution: Cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial volume and costs; and manpower staffing costs. Using equipment and consumables costs and inventory data supplied by the user, CECP calculates unit cost factors and then combines these factors with transportation and burial cost algorithms to produce a complete report of decommissioning costs. In addition to costs, CECP also calculates person-hours, crew-hours, and exposure person-hours associated with decommissioning. 3 - Restrictions on the complexity of the problem: The program is designed for a specific waste charge structure. The waste cost data structure cannot handle intermediate waste handlers or changes in the charge rate structures. The decommissioning of a reactor can be divided into 5 periods. 200 different items for special equipment costs are possible. The maximum amount for each special equipment item is 99,999,999$. You can support data for 10 buildings, 100 components each; ESTS1071/01: There are 65 components for 28 systems available to specify the contaminated systems costs (BWR). ESTS1071/02: There are 75 components for 25 systems available to specify the contaminated systems costs (PWR)

  10. Workers doses in central European PWR NPPs

    International Nuclear Information System (INIS)

    Janzekovic, H.; Krizman, M.

    2003-01-01

    As is stated, the ISOE database which was established in 1992 forms an excellent basis for studies and comparisons of occupational exposure data between nuclear power plants. In the year 2001, 69% of all participating reactors were pressurised water reactors. The ISOE database presents workers' exposure from 213 participating pressurised reactors (PWR) from 27 countries in that year. Among these 32 PWRs belong to six Central European Countries. The analysis of the exposure of workers based on radiation protection performance indicators (collective dose, average dose etc.) in these PWRs could be related to some nuclear safety performance indicators for recent years using ISOE database. The comparison is made to ISOE world - wide data. In the six Central European Countries altogether 32 PWR operated in the year 2001.The international databases of performance indicators related to radiation protection as for example the ISOE or the UNSCEAR database can be use as an efficient tool in the management of radiation protection of workers in a nuclear facilities and regulatory bodies. The databases enable the study of performance trends and the improvement of radiation protection. (authors)

  11. A complex study on the reliability assessment of the containment of a PWR. Part I - Magnitude and probability of internal load behavior

    International Nuclear Information System (INIS)

    Augustin, W.; Kafka, P.

    1977-01-01

    For evaluation of the reliability of the safety enclosure in the case of accidents the time-dependent loads by internal pressure and temperature on the spheric steel containment and the correspondent probabilities had to be calculated. Of the spectrum of possible accidents, e.g. large LOCA which leads to a maximum pressure of approximately 4.7 bar. working of all safety systems presumed, small LOCA or rupture of a primary steam pipe, only those have been selected which result in a considerable increase of internal pressure in the safety containment. The pressure buildup in the steel containment depends roughly on the radioactive decay energy produced in the containment, on the performance of the safety systems operative after the accident and on the energy absorbed and transferred by the structural parts of the containment. For simplification the analysis of system behavior was performed in separate steps. Analysis was started by evaluation of alternate possibilities of pressure buildup depending on the function of different safety systems. Then the time dependent changes of temperature and pressure in the containment were calculated as well as the probabilities of the occurrence of the different maximum pressures. Technical data and accident event sequences describing the system analysed were taken from the PWR Biblis B, which at this time is typical for the PWR-line construction in the FRG. In order to avoid event sequences leading to complicated physical phenomena such sequences were selected which allowed well-defined description of consequences as hydrogen production by reaction of water with the Zircalloy fuel cladding or pressure buildup by CO 2 or steam generated from concrete getting in contact with the core-melt. The computer code ZOCO VI was used to calculate pressure buildup for the different event sequences. This code calculates time dependence of pressure and temperature in a multiply segmented safety containment considering accumulation and

  12. Source terms associated with two severe accident sequences in a 900 MWe PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Berthion, Y.; Lhiaubet, G.; Lucas, M.

    1983-12-01

    Hypothetical accidents taken into account in PWR risk assessment result in fission product release from the fuel, transfer through the primary circuit, transfer into the reactor containment building (RCB) and finally release to the environment. The objective of this paper is to define the characteristics of the source term (noble gases, particles and volatile iodine forms) released from the reactor containment building during two dominant core-melt accident sequences: S 2 CD and TLB according to the ''Reactor Safety Study'' terminology. The reactor chosen for this study is a French 900 MWe PWR unit. The reactor building is a prestressed concrete containment with an internal liner. The first core-melt accident sequence is a 2-break loss-of-coolant accident on the cold leg, with failure of both system and the containment spray system. The second one is a transient initiated by a loss of offsite and onsite power supply and auxiliary feedwater system. These two sequences have been chosen because they are representative of risk dominant scenarios. Source terms associated with hypothetical core-melt accidents S 2 CD and TLB in a French PWR -900 MWe- have been performed using French computer codes (in particular, JERICHO Code for containment response analysis and AEROSOLS/31 for aerosol behavior in the containment)

  13. 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)

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

    International Nuclear Information System (INIS)

    Peryoga, Yoga; Saito, Masaki; Artisyuk, Vladimir

    2002-01-01

    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 241 Am might play the decisive role of burnable absorber to reduce the initial reactivity excess while the short-lived nuclides 242 Cm and 244 Cm decay into even plutonium isotopes, thus increasing the extent of denaturation for primary fissile 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 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% 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)

  15. PWR radiation fields at combustion engineering plants through mid-1985: Final report

    International Nuclear Information System (INIS)

    Barshay, S.S.; Beineke, T.A.; Bradshaw, R.W.

    1987-01-01

    This report presents the results of the initial phase of the EPRI-PWR Standard Radiation Monitoring Program (SRMP) for PWR nuclear power plants with Nuclear Steam Supply Systems supplied by Combustion Engineering, Inc. The purposes of the SRMP are to provide reliable, consistent and systematic measurements of the rate of radiation-field buildup at operating PWR's; and to use that information to identify opportunities for radiation control and the consequent reduction of occupational radiation exposure. The report includes radiation surveys from seven participating power plants. These surveys were conducted at well-defined locations on the reactor coolant loop piping and steam generators, and/or inside the steam generator channel heads. In most cases only one survey is available from each power plant, so that conclusions about the rate of radiation-field buildup are not possible. Some observations are made about the distribution pattern of radiation levels within the steam generator channel heads and around the reactor coolant loops. The report discusses the relationship between out-of-core radiation fields (as measured by the SRMP) and: the pH of the reactor coolant, the concentration of lithium hydroxide in the reactor coolant, and the frequency of changes in reactor power level. In order to provide data for possible future correlations of these parameters with the SRMP radiation-field data, the report summarizes information available from participating plants on primary coolant pH, and on the frequency of changes in reactor power level. 12 refs., 22 figs., 7 tabs

  16. Definition of thermal-hydraulics parameters of a naval PWR via energy balance of a Westinghouse PWR

    Energy Technology Data Exchange (ETDEWEB)

    Chaves, Luiz C.; Curi, Marcos F., E-mail: marcos.curi@cefet-rj.br [Centro Federal de Educação Tecnológica Celso Suckow da Fonseca (CEFET-RJ), Rio de Janeiro, RJ (Brazil). Department of Mechanical Engineering

    2017-07-01

    In this work, we used the operational parameters of the Angra 1 nuclear power plant, designed by Westinghouse, to estimate the thermal-hydraulic parameters for naval nuclear propulsion, focusing on the analysis of the reactor and steam generator. A thermodynamics analysis was made to reach the operational parameters of primary circuit such as pressure, temperature, power generated among others. Previous studies available in literature of 2-loop Westinghouse Nuclear Power Plants, which is based on a PWR and similar to Angra-1, support this analysis in the sense of a correct procedure to deal with many complex processes to energy generation from a nuclear source. Temperature profiles in reactor and steam generator were studied with concepts of heat transfer, fluid mechanics and also some concepts of nuclear systems, showing the behavior into them. In this simulation, the Angra 1 primary circuit was reduced on a scale of 1: 3.5 to fit in a Scorpène-class submarine. The reactor generates 85.7 MW of total thermal power. The maximum power and temperatures reached were lower than the operational safe limits established by Westinghouse. The number of tubes of the steam generator was determined in 990 U-tubes with 6.3 m of average length. (author)

  17. PWR system simulation and parameter estimation with neural networks

    International Nuclear Information System (INIS)

    Akkurt, Hatice; Colak, Uener

    2002-01-01

    A detailed nonlinear model for a typical PWR system has been considered for the development of simulation software. Each component in the system has been represented by appropriate differential equations. The SCILAB software was used for solving nonlinear equations to simulate steady-state and transient operational conditions. Overall system has been constructed by connecting individual components to each other. The validity of models for individual components and overall system has been verified. The system response against given transients have been analyzed. A neural network has been utilized to estimate system parameters during transients. Different transients have been imposed in training and prediction stages with neural networks. Reactor power and system reactivity during the transient event have been predicted by the neural network. Results show that neural networks estimations are in good agreement with the calculated response of the reactor system. The maximum errors are within ±0.254% for power and between -0.146 and 0.353% for reactivity prediction cases. Steam generator parameters, pressure and water level, are also successfully predicted by the neural network employed in this study. The noise imposed on the input parameters of the neural network deteriorates the power estimation capability whereas the reactivity estimation capability is not significantly affected

  18. PWR system simulation and parameter estimation with neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Akkurt, Hatice; Colak, Uener E-mail: uc@nuke.hacettepe.edu.tr

    2002-11-01

    A detailed nonlinear model for a typical PWR system has been considered for the development of simulation software. Each component in the system has been represented by appropriate differential equations. The SCILAB software was used for solving nonlinear equations to simulate steady-state and transient operational conditions. Overall system has been constructed by connecting individual components to each other. The validity of models for individual components and overall system has been verified. The system response against given transients have been analyzed. A neural network has been utilized to estimate system parameters during transients. Different transients have been imposed in training and prediction stages with neural networks. Reactor power and system reactivity during the transient event have been predicted by the neural network. Results show that neural networks estimations are in good agreement with the calculated response of the reactor system. The maximum errors are within {+-}0.254% for power and between -0.146 and 0.353% for reactivity prediction cases. Steam generator parameters, pressure and water level, are also successfully predicted by the neural network employed in this study. The noise imposed on the input parameters of the neural network deteriorates the power estimation capability whereas the reactivity estimation capability is not significantly affected.

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

    International Nuclear Information System (INIS)

    Bahn, C.B.; Han, B.C.; Bum, J.S.; Hwang, I.S.; Lee, C.B.

    2002-01-01

    Radioactive nuclei, such as 68 Co and 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), 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)

  20. PWR control rod ejection analysis with the numerical nuclear reactor

    International Nuclear Information System (INIS)

    Hursin, M.; Kochunas, B.; Downar, T. J.

    2008-01-01

    During the past several years, a comprehensive high fidelity reactor LWR core modeling capability has been developed and is referred to as the Numerical Nuclear Reactor (NNR). The NNR achieves high fidelity by integrating whole core neutron transport solution and ultra fine mesh computational fluid dynamics/heat transfer solution. The work described in this paper is a preliminary demonstration of the ability of NNR to provide a detailed intra pin power distribution during a control rod ejection accident. The motivation of the work is to quantify the impact on the fuel performance calculation of a more physically accurate representation of the power distribution within the fuel rod during the transient. The paper addresses first, the validation of the transient capability of the neutronic module of the NNR code system, DeCART. For this purpose, a 'mini core' problem consisting of a 3x3 array of typical PWR fuel assemblies is considered. The initial state of the 'mini core' is hot zero power with a control rod partially inserted into the central assembly which is fresh fuel and is adjacent to once and twice burned fuel representative of a realistic PWR arrangement. The thermal hydraulic feedbacks are provided by a simplified fluids and heat conduction solver consistent for both PARCS and DeCART. The control rod is ejected from the central assembly and the transient calculation is performed with DeCART and compared with the results of the U.S. NRC core simulation code PARCS. Because the pin power reconstruction in PARCS is based on steady state intra assembly pin power distributions which do not account for thermal feedback during the transient and which do not take into account neutron leakage from neighboring assemblies during the transient, there are some small differences in the PARCS and DeCART pin power prediction. Intra pin power density information obtained with DeCART represents new information not available with previous generation of methods. The paper then

  1. The simulation research for the dynamic performance of integrated PWR

    International Nuclear Information System (INIS)

    Yuan Jiandong; Xia Guoqing; Fu Mingyu

    2005-01-01

    The mathematical model of the reactor core of integrated PWR has been studied and simplified properly. With the lumped parameter method, authors have established the mathematical model of the reactor core, including the neutron dynamic equation, the feedback reactivities model and the thermo-hydraulic model of the reactor. Based on the above equations and models, the incremental transfer functions of the reactor core model have been built. By simulation experimentation, authors have compared the dynamic characteristics of the integrated PWR with the traditional dispersed PWR. The simulation results show that the mathematical models and equations are correct. (authors)

  2. A universal PWR spectral history correction

    International Nuclear Information System (INIS)

    Hutt, P.K.; Nunn, D.L.

    1989-01-01

    The accuracy of a form of universal correction for the difference between depletion conditions assumed in PWR assembly lattice calculations and those experienced in a reactor burn-up is investigated. The correction is based on lattice calculations in which only one such depletion history difference, depletion at two different water densities, is explicitly represented by lattice calculations. The assumption is made that other historical effects bear the same relationship to an appropriate time-average of the two-group neutron flux spectrum. The correction is shown to be accurate for the most important historical effects, depletion with burnable absorbers inserted, control rods inserted or at a different soluble boron level, in addition to density itself. The correction is less accurate for representing depletion at a different fuel or coolant temperature but even in these cases gives an improvement over no correction. In addition it is argued that these historic temperature effects are likely to be of minor importance. (author)

  3. Evaluation model for PWR irradiated fuel

    International Nuclear Information System (INIS)

    Gomes, I.C.

    1983-01-01

    The individual economic value of the plutonium isotopes for the recycle of the PWR reactor is investigated, assuming the existence of an market for this element. Two distinct market situations for the stages of the fuel cycle are analysed: one for the 1972 costs and the other for costs of 1982. Comparisons are made for each of the two market situations concerning enrichment of the U-235 in the uranium fuel that gives the minimum cost in the fuel cycle. The method adopted to establish the individual value of the plutonium isotopes consists on the economical analyses of the plutonium fuel cycle for four different isotopes mixtures refering to the uranium fuel cycle. (Author) [pt

  4. Chemical cleaning of nuclear (PWR) steam generators

    International Nuclear Information System (INIS)

    Welty, C.S. Jr.; Mundis, J.A.

    1982-01-01

    This paper reports on a significant research program sponsored by a group of utilities (the Steam Generator Owners Group), which was undertaken to develop a process to chemically remove corrosion product deposits from the secondary side of pressurized water reactor (PWR) power plant steam generators. Results of this work have defined a process (solvent system and application methods) that is capable of removing sludge and tube-to-tube support plate crevice corrosion products generated during operation with all-volatile treatment (AVT) water chemistry. Considers a plant-specific test program that includes all materials in the steam generator to be cleaned and accounts for the physical locations (proximity and contact) of those materials. Points out that prior to applying the process in an operational unit, the utility, with the participation of the NSSR vendor, must define allowable total corrosion to the materials of construction of the unit

  5. Stochastic optimization of loading pattern for PWR

    International Nuclear Information System (INIS)

    Smuc, T.; Pevec, D.

    1994-01-01

    The application of stochastic optimization methods in solving in-core fuel management problems is restrained by the need for a large number of proposed solutions loading patterns, if a high quality final solution is wanted. Proposed loading patterns have to be evaluated by core neutronics simulator, which can impose unrealistic computer time requirements. A new loading pattern optimization code Monte Carlo Loading Pattern Search has been developed by coupling the simulated annealing optimization algorithm with a fast one-and-a-half dimensional core depletion simulator. The structure of the optimization method provides more efficient performance and allows the user to empty precious experience in the search process, thus reducing the search space size. Hereinafter, we discuss the characteristics of the method and illustrate them on the results obtained by solving the PWR reload problem. (authors). 7 refs., 1 tab., 1 fig

  6. Evaluation of tight-pitch PWR cores

    International Nuclear Information System (INIS)

    Correa, F.; Driscoll, M.J.; Lanning, D.D.

    1979-08-01

    The impact of tight pinch cores on the consumption of natural uranium ore has been evaluated for two systems of coupled PWR's namely one particular type of thorium system - 235 U/UO 2 : Pu/ThO 2 : 233 U/ThO 2 - and the conventional recycle-mode uranium system - 235 U/UO 2 : Pu/UO 2 . The basic parameter varied was the fuel-to-moderator volume ratio (F/M) of the (uniform) lattice for the last core in each sequence. Although methods and data verification in the range of present interest, 0.5 (current lattices) 1.0, the EPRI-LEOPARD and LASER programs used for the thorium and uranium calculations, respectively, were successfully benchmarked against several of the more pertinent experiments

  7. 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)

    1999-12-31

    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)

  8. Zebra: An advanced PWR lattice code

    International Nuclear Information System (INIS)

    Cao, L.; Wu, H.; Zheng, Y.

    2012-01-01

    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)

  9. Full MOX high burn-up PWR

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu; Kugo, Teruhiko; Shimada, Shoichiro; Araya, Fumimasa; Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    As a part of conceptual investigation on advanced light water reactors for the future, a light water reactor with the high burn-up of 100 GWd/t, the long cycle operation of 3 years and the full MOX core is being studied, aiming at the improvement on economical aspects, the reduction of the spent fuel production, the utilization of Plutonium and so forth. The present report summarizes investigation on PWR-type reactors. The core with the increased moderation of the moderator-to-fuel volume ratio of 2.6 {approx} 3.0 has been proposed be such a core that accomplishes requirements mentioned above. Through the neutronic and the thermo-hydrodynamic evaluation, the performances of the core have been evaluated. Also, the safety designing is underway considering the reactor system with the passive safety features. (author)

  10. 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)

  11. 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)

  12. Development of advanced PWR steam generator

    International Nuclear Information System (INIS)

    Saito, Itaru; Nakamura, Tomomichi

    1999-01-01

    In response to the increased power of the advanced PWR, it is necessary to develop a steam generator (SG) which has a large capacity with high performance and high reliability as well as being economical to produce. In this paper, the development of the design of a new SG for the advanced PWRs is described and compared with the design of a conventional SG. Moreover, an outline of a seismic verification test for the U-bend tube bundle which includes advanced anti-vibration bars (AVB) which are very important is described. As a result, it was verified that the bundle has sufficient strength and a relatively high attenuation to seismic loads. These results will be reflected in the detailed design of advanced AVBs. (author)

  13. Radiation embrittlement of PWR vessel supports

    International Nuclear Information System (INIS)

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

    1989-01-01

    Several studies pertaining to radiation damage of PWR vessel supports were conducted between 1978 and 1987. During this period, apparently there was no reason to believe that low-temperature (<100 degree C) MTR embrittlement data were not appropriate for evaluating embrittlement of PWR vessel supports. However, late in 1986, data from the High Flux Isotope Reactor (HFIR) vessel surveillance program indicated that the embrittlement rates of the several HFIR vessel materials (A212-B, A350-LF3, A105-II) were substantially greater than anticipated on the basis of MTR data. Further evaluation of the HFIR data suggested that a fluence-rate effect was responsible for the apparent discrepancy, and shortly thereafter it became apparent that this rate effect was applicable to the evaluation of LWR vessel supports. As a result, the Nuclear Regulatory Commission (NRC) requested that the Oak Ridge National Laboratory (ORNL) evaluate the impact of the apparent embrittlement rate effect on the integrity of light-water-reactor (LWR) vessel supports. The purpose of the study was to provide an indication of whether the integrity of reactor vessel supports is likely to be challenged by radiation-induced embrittlement. The scope of the evaluation included correlation of the HFIR data for application to the evaluation of LWR vessel supports; a survey and cursory evaluation of all US LWR vessel support designs, selection of two plants for specific-plant evaluation, and a specific-plant evaluation of both plants to determine critical flaw sizes for their vessel supports. 19 refs., 8 figs., 2 tabs

  14. Aerosols behavior inside a PWR during an accident

    International Nuclear Information System (INIS)

    Hervouet, C.

    1983-01-01

    During very hypothetical accidents occurring in a pressurized water ractor, radioactive aerosols can be released, during core-melt, inside the reactor containment building. A good knowledge of their behavior in the humid containment atmosphere (mass concentration and size distribution) is essential in order to evaluate their harmfulness in case of environment contamination and to design possible filtration devices. Accordingly the Safety Analysis Department of the Atomic Energy Commission uses several computer models, describing the particle formation (BOIL/MARCH), then behavior in the primary circuits (TRAP-MELT), and in the reactor containment building (AEROSOLS-PARFDISEKO-III B). On the one hand, these models have been improved, in particular the one related to the aerosol formation (nature and mass of released particles) using recent experimental results. On the other hand, sensitivity analyses have been performed with the AEROSOLS code which emphasize the particle coagulation parameters: agglomerate shape factors and collision efficiency. Finally, the different computer models have been applied to the study of aerosol behavior during a 900 MWe PWR accident: loss-of-coolant-accident (small break with failure of all safety systems) [fr

  15. Structural integrity evaluation of PWR nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Cruz, Julio R.B.; Mattar Neto, Miguel

    1999-01-01

    The reactor pressure vessel (RPV) is the most important structural component of a PWR nuclear power plant. It contains the reactor core and is the main component of the primary system pressure boundary, the system responsible for removing the heat generated by the nuclear reactions. It is considered not replaceable and, therefore, its lifetime is a key element to define the plant life as a whole. Three critical issues related to the reliability of the RPV structural integrity come out by reason of the radiation damage imposed to the vessel material during operation. These issues concern the definition of pressure versus temperature limits for reactor heatup and cooldown, pressurized thermal shock evaluation and assessment of reactor vessels with low upper shelf Charpy impact energy levels. This work aims to present the major aspects related to these topics. The requirements for preventing fracture of the RPV are reviewed as well as the available technology for assessing the safety margins. For each mentioned problem, the several steps for structural integrity evaluation are described and the analysis methods are discussed. (author)

  16. Applicability of oxygenated water chemistry for PWR secondary systems

    Energy Technology Data Exchange (ETDEWEB)

    Hermansson, H.P. [Studsvik Nuclear AB, Nykoeping (Sweden); Takiguchi, H.; Otoha, K. [Japan Atomic Power Co., Tokyo (Japan)

    2002-07-01

    Introduction of oxygenated water chemistry (OWC) in PWR secondary side is considered as a means to reduce the transportation of corrosion products into the steam generator and thus also minimizing crevice deposits and subsequent materials problems. One main concern, however, is the risk of inter-granular attack (IGA) in crevices. In order to study effects on crevice tube IGA by OWC, a series of experiments were performed in a steam generator (SG) simulating loop. This comprised a SG tube and a tube support plate (TSP) together forming the crevice. The over-all objective of the work accounted here was to demonstrate that it is possible to operate the steam generator secondary side with OWC without causing intolerable IGA or other types of attack on the tube in the crevice area. Tubes of sensitized Alloy 600 were exposed during a total of nine experiments in an autoclave using a TSP/tube arrangement with an asymmetric crevice design. Experiments were performed at high and low pH and potential under open and packed crevice conditions. The aggressiveness of the crevice environment was also further increased by addition of carbonate and chloride. Furthermore the tube was pressurized. Experimental parameters were monitored on the primary side as well as in the secondary bulk phase and in the crevice. (authors)

  17. Effects of Nutrient Dynamics, Light and Temperature on the Patchiness of Phytoplankton and Primary Production in the Estuarine and Coastal Zones of Liaodong Bay, China: A Typical Case Study

    Science.gov (United States)

    Pei, S.; Laws, E. A.; Ye, S.

    2017-12-01

    Fluvial inputs of nutrients and efficient nutrient recycling mechanisms make estuarine and coastal zones highly productive bodies of water. For the same reasons, they are susceptible to eutrophication problems. In China, eutrophication problems along coasts are becoming serious because of discharges of domestic sewage and industrial wastewater and runoff of agricultural fertilizer. Addressing these problems requires an informed assessment of the factors that controlling algal production. Our study aims at determining the factors that controlling patchiness of phytoplankton and primary production in Liaodong Bay, China that receives large inputs of nutrients from human activities in its watershed, and examining the variation patterns of phytoplankton photosynthesis under both stressors of climate change and human activities. Results of our field study suggest that nutrient concentrations were above growth-rate-saturating concentrations throughout Liaodong bay, with the possible exception of phosphate at some stations. This assessment was consistent with the results of nutrient enrichment experiments and the values of light-saturated photosynthetic rates and areal photosynthetic rates. Two large patches of high biomass and production with dimensions on the order of 10 km reflect the effects of water temperature and variation of light penetration restricted by water turbidity. To examine the effects of irradiance and temperature on light-saturated photosynthetic rates normalized to chlorophyll a concentrations (Popt), light-conditioned Popt values were modeled as a function of the temperature with a satisfactory fit to our field data (R2 = 0.60, p = 0.003). In this model, light-conditioned Popt values increased with temperatures from 22°C to roughly 25°C but declined precipitously at higher temperatures. The relatively high Popt values and low ratios of light absorbed to photosynthesis at coastal stations suggest the highly efficient usage of absorbed light by

  18. Monte Carlo Simulation of Quantitative Electron Probe Microanalysis of the PWR Spent Fuel with a Pt Coating

    International Nuclear Information System (INIS)

    Kwon, Hyoung Mun; Lee, Hyung Kwon; Son, Young Zoon; Chun, Yong Bum

    2012-01-01

    The PWR spent fuel sample should be coated with conducting material in order to provide a path for electrons and to prevent charging. Generally, the ZAF method has been used for quantitative electron probe microanalysis of conducting samples. However, the coated samples are not applicable for the ZAF method. Probe current, primary electron energy and x-ray produced by the primary beam are attenuated within the coating films. The electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program [2] to evaluate the x-ray attenuation within the Pt coating films. The target samples are the PWR spent fuels with 50 GWd/tU of burnup , 6 years of cooling time and a Pt coating film (3, 5, 7, 10 and 15 nm thickness)

  19. Monte Carlo Simulation of Quantitative Electron Probe Microanalysis of the PWR Spent Fuel with a Pt Coating

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyoung Mun; Lee, Hyung Kwon; Son, Young Zoon; Chun, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    The PWR spent fuel sample should be coated with conducting material in order to provide a path for electrons and to prevent charging. Generally, the ZAF method has been used for quantitative electron probe microanalysis of conducting samples. However, the coated samples are not applicable for the ZAF method. Probe current, primary electron energy and x-ray produced by the primary beam are attenuated within the coating films. The electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program [2] to evaluate the x-ray attenuation within the Pt coating films. The target samples are the PWR spent fuels with 50 GWd/tU of burnup , 6 years of cooling time and a Pt coating film (3, 5, 7, 10 and 15 nm thickness)

  20. Liquid radioactive waste processing improvement of PWR nuclear power plants; Melhorias no processamento de rejeitos liquidos radioativos de usinas nucleares PWR

    Energy Technology Data Exchange (ETDEWEB)

    Nery, Renata Wolter dos Reis; Martinez, Aquilino Senra; Monteiro, Jose Luiz Fontes [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear]. E-mail: wolter@eletronuclear.gov.br; monteiro@peq.coppe.ufrj.br; aquilinosenra@lmp.ufrj.br

    2005-07-01

    The study evaluate an inorganic ion exchange to process the low level liquid radwaste of PWR nuclear plants, so that the level of the radioactivity in the effluents and the solid waste produced during the treatment of these liquid radwaste can be reduced. The work compares two types of ion exchange materials, a strong acid cation exchange resin, that is the material typically used to remove radionuclides from PWR nuclear plants wastes, and a mordenite zeolite. These exchange material were used to remove cesium from a synthetic effluent containing only this ion and another effluent containing cesium and cobalt. The breakthrough curves of the zeolite and resin using a fix bed reactor were compared. The results demonstrated that the zeolite is more efficient than the resin in removing cesium from a solution containing cesium and cobalt. The results also showed that a bed combining zeolite and resin can process more volume of an effluent containing cesium and cobalt than a bed resin alone. (author)

  1. Testing typicality in multiverse cosmology

    Science.gov (United States)

    Azhar, Feraz

    2015-05-01

    In extracting predictions from theories that describe a multiverse, we face the difficulty that we must assess probability distributions over possible observations prescribed not just by an underlying theory, but by a theory together with a conditionalization scheme that allows for (anthropic) selection effects. This means we usually need to compare distributions that are consistent with a broad range of possible observations with actual experimental data. One controversial means of making this comparison is by invoking the "principle of mediocrity": that is, the principle that we are typical of the reference class implicit in the conjunction of the theory and the conditionalization scheme. In this paper, we quantitatively assess the principle of mediocrity in a range of cosmological settings, employing "xerographic distributions" to impose a variety of assumptions regarding typicality. We find that for a fixed theory, the assumption that we are typical gives rise to higher likelihoods for our observations. If, however, one allows both the underlying theory and the assumption of typicality to vary, then the assumption of typicality does not always provide the highest likelihoods. Interpreted from a Bayesian perspective, these results support the claim that when one has the freedom to consider different combinations of theories and xerographic distributions (or different "frameworks"), one should favor the framework that has the highest posterior probability; and then from this framework one can infer, in particular, how typical we are. In this way, the invocation of the principle of mediocrity is more questionable than has been recently claimed.

  2. 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.

  3. Simulation of the fuel rod thermal hydraulic performance during the blow down phase in a PWR

    International Nuclear Information System (INIS)

    Gadelha, J.A.M.

    1982-10-01

    A digital computer code to predict the fuel rod thermalhydraulic performance during a postulated loss-of-coolant accident (LOCA) in the primary circuit of a PWR nuclear power plant is developed. The fuel rod corresponds to that in an average channel in the core. Only the blowdown phase is considered during the accident. The conservation equations of mass, momentum, and energy, and the heat conduction equation are solved to determine the fuel rod conditions during the accident. Finite differences are applied as a numerical method in the solution of the equations modelling the rod and coolant conditions. (Author) [pt

  4. Long-term preventive maintenance of instrumentation control equipment for PWR plants

    International Nuclear Information System (INIS)

    Sugitani, S.; Nanba, M.

    2006-01-01

    Since the PWR plants in Japan have been operated more than 30 years, main instrumentation control equipment of analog systems has been renewed to digital control systems. Renewal works had to be done in short period within periodical inspection term and for several facilities. The Mitsubishi LTD group had been provided with these market needs by its digital control system (MELTAC-NplusR 3) applicable to main instrumentation control equipment for primary and secondary systems and had already finished the renewal for practical plants. (T. Tanaka)

  5. Steam generator tube rupture risk impact on design and operation of French PWR plants

    International Nuclear Information System (INIS)

    Depond, G.; Sureau, H.

    1984-01-01

    The experience of steam generator tube leaks incidents in PWR plants has resulted in an increase of EDF analysis leading to improvements in design and post-accidental operation for new projects and operating plants. The accident consequences are minimized for each of the NSSS three barriers: first barrier: safeguard systems design and operating procedures relying upon core safety allow to maintain a low level of primary radioactivity, second barrier: steam generator design and periodic inspection allow to reduce tube ruptures risks and third barrier: atmospheric releases are reduced as a result of optimal recovery procedures, detection improvements and atmospheric steam valves design improvements. (orig.)

  6. Decision tree based knowledge acquisition and failure diagnosis using a PWR loop vibration model

    International Nuclear Information System (INIS)

    Bauernfeind, V.; Ding, Y.

    1993-01-01

    An analytical vibration model of the primary system of a 1300 MW PWR was used for simulating mechanical faults. Deviations in the calculated power density spectra and coherence functions are determined and classified. The decision tree technique is then used for a personal computer supported knowledge presentation and for optimizing the logical relationships between the simulated faults and the observed symptoms. The optimized decision tree forms the knowledge base and can be used to diagnose known cases as well as to include new data into the knowledge base if new faults occur. (author)

  7. Consideration of loading conditions initiated by thermal transients in PWR pressure vessels

    International Nuclear Information System (INIS)

    Azodi; Glahn; Kersting; Schulz; Jansky.

    1983-01-01

    This report describes the present state of PWR-plants in the Federal Republic of Germany with respect to - the design of the primary pressure boundary - the analysis of thermal transients and resulting loads - the material conditions and neutron fluence - the requirements for protection against fast fracture. The experimental and analytical research and development programs are delineated together with some foreign R and D programs. It is shown that the parameters investigated (loading condition, crack shape and orientation etc.) cover a broad range. Extensive analytical investigations are emphasized. (orig./RW) [de

  8. Design of a steam generator for PWR power plants and steady state simulation

    International Nuclear Information System (INIS)

    Ferreira, W.J.

    1982-01-01

    A procedure and a computer code for the thermal design of a steam generator for PWR power plants is developed. A vertical integral steam generator with inverted U-tubes and natural circulation of the secondary side is selected for modelling. Primary fluid velocity and recirculation ratio are varied to obtain the preliminary dimensions. Further, adjustments are made through iteractive solution of the equations of conservation of mass, energy and momentum. An agreement is found between design calculations for steam generators of different capacities and existing designs. (Author) [pt

  9. Cyclic elastic analysis of a PWR nozzle subjected to a repeated thermal shock

    International Nuclear Information System (INIS)

    Locci, J.M.; Prost, J.P.

    1979-01-01

    In the primary piping system of a PWR nuclear power plant, some nozzles are subjected to strong thermal shocks due to sudden thermal variations in the internal water flow. The thermal gradients are sufficiently high to induce general elastic plastic behaviour. The design of these nozzles using the simplified elastic plastic analysis given in the ASME III Code NB-3200 generally leads to a very high usage factor. The aim of this work is to show by giving an example that a complete cyclic elastic plastic analysis makes it possible to considerably reduce the usage factor. (orig.)

  10. Behavior of a nine-rod PWR bundle under power-cooling-mismatch conditions

    International Nuclear Information System (INIS)

    Gunnerson, F.S.; Sparks, D.T.

    1979-01-01

    An experiment to characterize the behavior of a nine-rod pressurized water reactor (PWR) fuel bundle operating during power-cooling-mismatch (PCM) conditions has been conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory (INEL). The experiment, designated Test PCM-5, is part of a series of PCM experiments designed to evaluate light water reactor (LWR) fuel rod response under postulated accident conditions. Test PCM-5 was the first nine-rod bundle experiment in the PCM test series. The primary objectives and the results of the experiment are described

  11. Changes in 900 MW PWR alarm processing policy

    Energy Technology Data Exchange (ETDEWEB)

    Pont, M [Electricite de France, Generation and Transmission, Nuclear Power Plant Operations, Paris (France)

    1997-09-01

    Following a brief description of the current 900 MW PWR alarm processing system, this document presents the feasibility study carried out within the scope of the Instrumentation and Control Refurbishment project (R2C). (author). 4 figs, tabs.

  12. Changes in 900 MW PWR alarm processing policy

    International Nuclear Information System (INIS)

    Pont, M.

    1997-01-01

    Following a brief description of the current 900 MW PWR alarm processing system, this document presents the feasibility study carried out within the scope of the Instrumentation and Control Refurbishment project (R2C). (author). 4 figs, tabs

  13. Deboration in nuclear stations of the PWR type

    International Nuclear Information System (INIS)

    1978-01-01

    Reactivity control in nuclear power stations of the PWR type is realised with boric acid. A method to concentrate boric acid without an evaporator has been studied. A flow-sheet with reverse osmosis is proposed. (author)

  14. Severe accident considerations for modern KWU-PWR plants

    International Nuclear Information System (INIS)

    Eyink, J.

    1987-01-01

    In assumption of severe accident on modern KWU-PWR plants the author discusses on the: selection of core meltdown sequences, course of the accident, containment behaviour and source terms for fission products release to the environment

  15. Dose rate evaluation after accident in a PWR

    International Nuclear Information System (INIS)

    Cladel, C.; Duchemin, B.; Le Dieu de Ville, A.; Nimal, B.; Nimal, J.C.; Evrard, J.M.

    1983-05-01

    A calculation scheme for the gamma radiation dose rate after accident in a PWR is presented. These studies use a fine description of the geometry and of the fission product inventory. Some results are given and some improvements are planned

  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. Ultrasonic inspection for testing the PWR fuel rod endplug welds

    International Nuclear Information System (INIS)

    Pillet, C.; Destribats, M.T.; Papezyk, F.

    1976-01-01

    A method of ultrasonic testing with local immersion and transversal waves was developed. It is possible to detect defects as the lacks of fusion and penetration and porosity in the PWR fuel rod endplug welds [fr

  18. Low-density moderation in the storage of PWR fuel assemblies

    International Nuclear Information System (INIS)

    Alcorn, F.M.

    1987-01-01

    The nuclear criticality safety of PWR fuel storage arrays requires that the potential of low-density moderation within the array be considered. The calculated criticality effect of low-density moderation in a typical PWR fuel assembly array is described in this paper. Calculated reactivity due to low-density moderation can vary significantly between physics codes that have been validated for well moderated systems. The availability of appropriate benchmark experiments for low-density moderation is quite limited; attempts to validate against the one set of suitable experiments at low density have been disappointing. Calculations indicate that a typical array may be unacceptable should the array be subjected to interstitial moderation equivalent to 5 % of full density water. Array parameters (such as spacing and size) will dramatically affect the calculated maximum K-eff at low-density moderation. Administrative and engineered control may be necessary to assure maintenance of safety at low-density moderation. Potential sources for low-density moderation are discussed; in general, accidentally achieving degrees of low-density moderation which might lead to a compromise of safety are not credible. (author)

  19. Assessment of fission product release from the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Generino, G.

    1984-07-01

    Fission product releases from the RCB associated with hypothetical core-melt accidents ABβ, S 2 CDβ and TLBβ in a PWR-900 MWe have been performed using French computer codes (in particular, the JERICHO Code for containment response analysis and AEROSOLS/B1 for aerosol behavior in the containment) related to thermalhydraulics and fission product behavior in the primary system and in the reactor containment building

  20. Model for transient simulation in a PWR steam circuit

    International Nuclear Information System (INIS)

    Mello, L.A. de.

    1982-11-01

    A computer code (SURF) was developed and used to simulate pressure losses along the tubes of the main steam circuit of a PWR nuclear power plant, and the steam flow through relief and safety valves when pressure reactors its thresholds values. A thermodynamic model of turbines (high and low pressure), and its associated components are simulated too. The SURF computer code was coupled to the GEVAP computer code, complementing the simulation of a PWR nuclear power plant main steam circuit. (Author) [pt

  1. GO evaluation of a PWR spray system. Final report

    International Nuclear Information System (INIS)

    Long, W.T.

    1975-08-01

    GO is a reliability analysis methodology developed over the years from 1960 to the present by Kaman Sciences Corporation, Colorado Springs, Colorado. In this report the GO methodology is presented and its application demonstrated by performing a reliability analysis of a conceptual PWR Containment Spray System. Certain numerical results obtained are compared with those of a prior fault tree analysis of the same system as documented in the 11 January 1973 draft report, A Fault Tree Evaluation of a PWR Spray System

  2. Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-1 and TSE-2

    International Nuclear Information System (INIS)

    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

  3. Experimental study of single- and two-phase flow fields around PWR steam generator tube support plates

    International Nuclear Information System (INIS)

    Bates, J.M.; Stewart, C.W.

    1979-08-01

    Laser-Doppler anemometry (LDA) was used to measure local mean axial velocities and turbulence intnsities at selected locations within a study model dimensionally protypic of an existing PWR steam generator design. The model tube bundle with support plate was installed in a special flow housing that formed part of an isothermal recirculating water flow loop. Flow conditions for this experiment were intended to simulate only typical single-phase flow velocities and were not an attempt to completely model actual steam generator, boiling, two-phase flow conditions. The measurements were performed in water at approximately 85 0 F with test section average velocities of approximately 0.55 and 1.1 fps. These conditions corresponded to Reynolds numbers of approximately 7,000 and approximately 14,000, respectively. Normalized velocity and turbulence intensity ratios are graphically reported. Additional qualitative, photographic investigations of air-water two-phase flows in a PWR steam generator study model were also performed

  4. Sample problem calculations related to two-phase flow transients in a PWR relief-piping network

    International Nuclear Information System (INIS)

    Shin, Y.W.; Wiedermann, A.H.

    1981-03-01

    Two sample problems related with the fast transients of water/steam flow in the relief line of a PWR pressurizer were calculated with a network-flow analysis computer code STAC (System Transient-Flow Analysis Code). The sample problems were supplied by EPRI and are designed to test computer codes or computational methods to determine whether they have the basic capability to handle the important flow features present in a typical relief line of a PWR pressurizer. It was found necessary to implement into the STAC code a number of additional boundary conditions in order to calculate the sample problems. This includes the dynamics of the fluid interface that is treated as a moving boundary. This report describes the methodologies adopted for handling the newly implemented boundary conditions and the computational results of the two sample problems. In order to demonstrate the accuracies achieved in the STAC code results, analytical solutions are also obtained and used as a basis for comparison

  5. Seismic analysis of a PWR 900 reactor: study of reactor building with soil-structure interaction and evaluation of floor spectra

    International Nuclear Information System (INIS)

    Gantenbein, F.; Aguilar, J.

    1983-08-01

    The purpose of this paper is the evaluation of seismic response and floor spectra for a typical PWR 900 reactor building with respect to soil-structure interaction for soil stiffness). The typical PWR 900 reactor building consists of a concrete cylindrical external building and roof dome, a concrete internal structure (internals) on a common foundation mat as illustrated. The seismic response is obtained by SRSS method and floor spectra directly from ground spectrum and modal properties of the structure. Seismic responses and floor spectra computation is performed in the case of two different ground spectra: EDF spectrum (mean of oscillator spectra obtained from 8 californian records) normalized to 0.2 g, and DSN spectrum (typical of shallow seism) normalized to 0.3 g. The first section is devoted to internals' modelisation, the second one to the axisymmetric model of the reactor, the third one to the seismic response, the fourth one to floor spectra

  6. Aging effects in PWR power plants components

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Diogo da S.; Guimaraes, Antonio C.F.; Moreira, Maria de Lourdes, E-mail: diogosb@outlook.com, E-mail: tony@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    This paper presents a contribution to the study of aging process of components in commercial plants of Pressurized Water Reactors (PWRs). The analysis is made through application of the Fault Trees Method, Monte Carlo Method and Fussell-Vesely Importance Measure. The approach of the study of aging in nuclear power plants, besides giving attention to the economic factors involved directly with the extent of their operational life, also provide significant data on security issues. The latest case involving process of life extension of a PWR could be seen in Angra 1 Nuclear Power Plant through investing of $27 million for the installation of a new reactor lid. The corrective action has generated an estimated operating life extension of Angra I in twenty years, offering great economy compared with building cost of a new plant and anterior decommissioning, if it had reached the time operating limit of forty years. The Extension of the operating life of a nuclear power plant must be accompanied by a special attention to the components of the systems and their aging process. After the application of the methodology (aging analysis of the injection system of the containment spray) proposed in this work, it can be seen that 'the increase in the rate of component failure, due the aging process, generates the increase in the general unavailability of the system that containing these basic components'. The final results obtained were as expected and may contribute to the maintenance policy, preventing premature aging process in Nuclear Plant Systems. (author)

  7. Alloy development for high burnup cladding (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-04-01

    An overview on current alloy development for high burnup PWR fuel cladding is given. It is mainly based on literature data. First, the reasons for an increase of the current mean discharge burnup from 35 MWd / kg(U) to 70 MWd / kg(U) are outlined. From the material data, it is shown that a batch average burnup of 60-70 MWd / kg(U), as aimed by many fuel vendors, can not be achieved with stand (=ASTM-) Zry-4 cladding tubes without violating accepted design criteria. Specifically criteria which limit maximum oxide scale thickness and maximum hydrogen content, and to a less degree, maximum creep and growth rate, can not be achieved. The development potential of standard Zry-4 is shown. Even when taking advantage of this potential, it is shown that an 'improved' Zry-4 is reaching its limits when it achieves the target burnup. The behavior of some Zr alloys outside the ASTM range is shown, and the advantages and disadvantages of the 3 alloy groups (ZrSn+transition metals, ZrNb, ZrSnNb+transition metals) which are currently considered to have the development potential for high burnup cladding materials are depicted. Finally, conclusions are drawn. (author). 14 refs., 11 tabs., 82 figs.

  8. 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.

  9. MELCOR/VISOR PWR desktop simulator

    International Nuclear Information System (INIS)

    With, Anka de; Wakker, Pieter

    2010-01-01

    Increasingly, there is a need for a learning support and training tool for nuclear engineers, utilities and students in order to broaden their understanding of advanced nuclear plant characteristics, dynamics, transients and safety features. Nuclear system analysis codes like ASTEC, RELAP5, RETRAN and MELCOR provide calculation results of and visualization tools can be used to graphically represent these results. However, for an efficient education and training a more interactive tool such as a simulator is needed. The simulator connects the graphical tool with the calculation tool in an interactive manner. A small number of desktop simulators exist [1-3]. The existing simulators are capable of representing different types of power plants and various accident conditions. However, they were found to be too general to be used as a reliable plant-specific accident analysis or training tool. A desktop simulator of the Pressurized Water Reactor (PWR) has been created under contract of the Dutch nuclear regulatory body (KFD). The desktop simulator is a software package that provides a close to real simulation of the Dutch nuclear power plant Borssele (KCB) and is used for training of the accident response. The simulator includes the majority of the power plant systems, necessary for the successful simulation of the KCB plant during normal operation, malfunctions and accident situations, and it has been successfully validated against the results of the safety evaluations from the KCB safety report. (orig.)

  10. Summary of PWR leak detection studies

    International Nuclear Information System (INIS)

    Cho, J.H.; Elia, F.A. Jr.

    1986-01-01

    Thermal-hydraulic analysis can be used to determine the location and magnitude of leaks inside and location of leaks outside a pressurized water reactor (PWR) containment as required by plant technical specifications. The major advantage of this detection method is that it minimizes radiation exposure of maintenance personnel because most of the leak detection process is performed from the control room outside containment. Plant-specific analyses are utilized to predict change in parameters such as local dew point temperature, relative humidity, dry bulb temperature, and flow rate to sump for various leak rates and enthalpies. These parameter responses are then programmed into the plant computer and instrumentation is provided for area monitoring. The actual inputs are continuously monitored and compared to the predicted plant responses to identify the leak location and quantify the leak. This study concludes that a system that monitors dew point (or relative humidity) and dry bulb temperature changes together with the flow rate to the sump will provide the capability to both locate and quantify a leak inside a containment, while a system that monitors dew point temperature (or relative humidity) changes will provide the capability to locate a leak outside a containment

  11. Hydrogen production in a PWR during LOCA

    International Nuclear Information System (INIS)

    Cassette, P.

    1984-01-01

    Hydrogen generation during a PWR LOCA has been estimated for design basis accident and for two more severe hypothetical accidents. Hydrogen production during design basis accident is a rather slow mechanism, allowing in the worst case, 15 days to connect a hydrogen recombining unit to the containment atmosphere monitoring system. Hydrogen generated by steam oxidation during more severe hypothetical accidents was found limited by steam availability and fuel melting phenomena. Uncertainty is, however, still remaining on corium-zirconium-steam interaction. In the worst case, calculations lead to the production of 500 kg of hydrogen, thus leading to a volume concentration of 15% in containment atmosphere, assuming homogeneous hydrogen distribution within the reactor building. This concentration is within flammability limits but not within detonation limits. However, hydrogen detonation due to local hydrogen accumulation cannot be discarded. A major uncertainty subsisting on hydrogen hazard is hydrogen distribution during the first hours of the accident. This point determines the effects and consequences of local detonation or deflagration which could possibly be harmful to safeguard systems, or induce missile generation in the reactor building. As electrical supply failures are identified as an important contributor to severe accident risk, corrective actions have been taken in France to improve their reliability, including the installation of a gas turbine on each site to supplement the existing sources. These actions are thus contributing to hydrogen hazard reduction

  12. Analysis of reactivity accidents in PWR'S

    International Nuclear Information System (INIS)

    Camous, F.; Chesnel, A.

    1989-12-01

    This note describes the French strategy which has consisted, firstly, in examining all the accidents presented in the PWR unit safety reports in order to determine for each parameter the impact on accident consequences of varying the parameter considered, secondly in analyzing the provisions taken into account to restrict variation of this parameter to within an acceptable range and thirdly, in checking that the reliability of these provisions is compatible with the potential consequences of transgression of the authorized limits. Taking into consideration violations of technical operating specifications and/or non-observance of operating procedures, equipment failures, and partial or total unavailability of safety systems, these studies have shown that fuel mechanical strength limits can be reached but that the probability of occurrence of the corresponding events places them in the residual risk field and that it must, in fact, be remembered that there is a wide margin between the design basis accidents and accidents resulting in fuel destruction. However, during the coming year, we still have to analyze scenarios dealing with cumulated events or incidents leading to a reactivity accident. This program will be mainly concerned with the impact of the cases examined relating to dilution incidents under normal operating conditions or accident operating conditions

  13. Conceptual study of advanced PWR core design

    International Nuclear Information System (INIS)

    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

  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. Multicriteria analysis of public protection in PWR's

    International Nuclear Information System (INIS)

    Lombard, J.

    1986-09-01

    In order to manage a risk efficiently and to reach the ALARA level of protection, the best possible protection options must be employed. As the available resources are limited, it is not always possible to choose those options that minimize the risk, therefore a compromise must be made between risks and safety expenses. When the choice is difficult or complex, finding such a compromise can be facilitated by resorting to a decision aiding method which allows the assessment of the respective advantages of the various protection options considered. The multicriteria methods employ successive comparisons. Instead of searching for a final indicator expressing the performance of each option they compare all option pairs in order to determine if the gap between their respective advantages and disadvantages is sufficient to estimate that one option of the each pair is better than the other. Instead of judging each option globally these methods evaluate the advantages and disadvantages associated with the eventual choice of an option as compared with the others. These differential and comparative approach gives more flexibility and allows the introduction of qualitative criteria. The method presented here (Electre 3), one of the most recent ones, allows a multicriteria comparison of a set of options keeping into account the uncertainties associated with the options or the preferences. In order to illustrate this method a simple example (4 options, 4 criteria) dealing with a PWR liquid releases treatment system, is taken up

  16. Modeling of PWR fuel at extended burnup

    Energy Technology Data Exchange (ETDEWEB)

    Dias, Raphael M.; Silva, Antonio Teixeira, E-mail: rmdias@ipen.br, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    Since FRAPCON-3 series was rolled out, many improvements have been implanted in fuel performance codes, based on most recent literature, to promote better predictions against current data. Much of this advances include: improving fuel gas release prediction, hydrogen pickup model, cladding corrosion, and many others. An example of those modifications has been new cladding materials has added into hydrogen pickup model to support M5™, ZIRLO™, and ZIRLO™ optimized family under pressurized water reactor (PWR) conditions. Recently some research have been made over USNRC's steady-state fuel performance code, assessments against FUMEX-III's data have concluded that FRAPCON provides best-estimate calculation of fuel performance. Face of this, a study is required to summarize all those modifications and new implementations, as well as to compare this result against FRAPCON's older version, scrutinizing FRAPCON-3 series documentation to understand the real goal and literature base of any improvements. We have concluded that FRAPCON's latest modifications are based on strong literature review. Those modifications were tested against most recent data to assure these results will be the best evaluation as possible. Many improvements have been made to allow USNRC to have an audit tool with the last improvements. (author)

  17. Operation and maintenance in Genkai PWR Plant

    International Nuclear Information System (INIS)

    Ohta, Shojiro

    1984-01-01

    The No.1 PWR plant with 559 MW capacity in the Genkai Nuclear Power Station, Kyushu Electric Power Co., Inc., required about 115 days for the regular inspection in fiscal 1982 and thereafter, although more maintenance work was done. But No.2 plant of the same type required not more than 80 days. In most cases, the period of one operation cycle was from 10 to 12 months, but in the third operation cycle of No.2 plant, it is expected to be 13 months. The capacity ratio of the whole power station was 75.2% at the end of fiscal 1983. These operational records all exceeded the Japanese average. The plants are two-loop Westinghouse type PWRs, and No.1 plant started the commercial operation of anti h and the increment of P 0 + . (author) apacity ratio of No.1 plant was 71.6%, and that of No.2 plant was 85.5%. The intergranular attack on steam generator tubes was found first in the fifth regular inspection, and also in the sixth and seventh inspections, and the faulty tubes were plugged. The prevention of its spread is the largest problem. The in-service quality assurance activity, the personnel training program and the effort of upgrading the plant availability are reported. (Kako, I.)

  18. Method of starting up PWR type reactor

    International Nuclear Information System (INIS)

    Kadokami, Akira; Ueno, Ryuji; Tsuge, Ayao; Onimura, Kichiro; Ochi, Tatsuya.

    1988-01-01

    Purpose: To start-up a PWR type reactor so as to effectively impregnate and concentrate corrosion inhibitors in intergranular corrosive faces. Method: Upon reactor start-up, after transferring from the warm zero output state to thermal power loaded state and injecting corrosion inhibitors, thermal power is returned to zero and, subsequently, increased up to a rated power. By selecting the thermal power upon injecting the corrosion inhibitors to a steam generator body, that is, by selecting a thermal power load that starts to boil in heat conduction tubes, feedwater in the clavis portion can be formed into an appropriate boiling convection and, accordingly, the corrosion inhibitors can be penetrated to the clevis portion at a higher rate and in a greater amount as compared with those under zero power condition. Subsequently, when the thermal power is reduced, a sub-cooled state is attained in the clevis portion, in which steams present in the intergranular corrosion faces in the heat conduction tubes are condensated. As a result, the corrosion inhibitors at high concentration are impregnated into the intergranular corrosive faces to provide excellent effects. (Kamimura, M.)

  19. Conceptual study on advanced PWR system

    International Nuclear Information System (INIS)

    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

  20. Computer aided information system for a PWR

    International Nuclear Information System (INIS)

    Vaidian, T.A.; Karmakar, G.; Rajagopal, R.; Shankar, V.; Patil, R.K.

    1994-01-01

    The computer aided information system (CAIS) is designed with a view to improve the performance of the operator. CAIS assists the plant operator in an advisory and support role, thereby reducing the workload level and potential human errors. The CAIS as explained here has been designed for a PWR type KLT- 40 used in Floating Nuclear Power Stations (FNPS). However the underlying philosophy evolved in designing the CAIS can be suitably adopted for other type of nuclear power plants too (BWR, PHWR). Operator information is divided into three broad categories: a) continuously available information b) automatically available information and c) on demand information. Two in number touch screens are provided on the main control panel. One is earmarked for continuously available information and the other is dedicated for automatically available information. Both the screens can be used at the operator's discretion for on-demand information. Automatically available information screen overrides the on-demand information screens. In addition to the above, CAIS has the features of event sequence recording, disturbance recording and information documentation. CAIS design ensures that the operator is not overburdened with excess and unnecessary information, but at the same time adequate and well formatted information is available. (author). 5 refs., 4 figs

  1. Modeling of PWR fuel at extended burnup

    International Nuclear Information System (INIS)

    Dias, Raphael M.; Silva, Antonio Teixeira

    2015-01-01

    Since FRAPCON-3 series was rolled out, many improvements have been implanted in fuel performance codes, based on most recent literature, to promote better predictions against current data. Much of this advances include: improving fuel gas release prediction, hydrogen pickup model, cladding corrosion, and many others. An example of those modifications has been new cladding materials has added into hydrogen pickup model to support M5™, ZIRLO™, and ZIRLO™ optimized family under pressurized water reactor (PWR) conditions. Recently some research have been made over USNRC's steady-state fuel performance code, assessments against FUMEX-III's data have concluded that FRAPCON provides best-estimate calculation of fuel performance. Face of this, a study is required to summarize all those modifications and new implementations, as well as to compare this result against FRAPCON's older version, scrutinizing FRAPCON-3 series documentation to understand the real goal and literature base of any improvements. We have concluded that FRAPCON's latest modifications are based on strong literature review. Those modifications were tested against most recent data to assure these results will be the best evaluation as possible. Many improvements have been made to allow USNRC to have an audit tool with the last improvements. (author)

  2. ABB PWR fuel design for high burnup

    International Nuclear Information System (INIS)

    Nilsson, S.; Jourdain, P.; Limback, M.; Garde, A.M.

    1998-01-01

    Corrosion, hydriding and irradiation induced growth of a based materials are important factors for the high burnup performance of PWR fuel. ABB has developed a number of Zr based alloys to meet the need for fuel that enables operation to elevated burnups. The materials include composition and processing optimised Zircaloy 4 (OPTIN TM ) and Zircaloy 2 (Zircaloy 2P), as well as advanced Zr based alloys with chemical compositions outside the composition specified for Zircaloy. The advanced alloys are either used as Duplex or as single component claddings. The Duplex claddings have an inner component of Zircaloy and an outer layer of Zr with small additions of alloying elements. ABB has furthermore improved the dimensional stability of the fuel assembly by developing stiffer and more bow resistant guide tubes while debris related fuel failures have been eliminated from ABB fuel by introducing the Guardian TM grid. Intermediate flow mixers that improve the thermal hydraulic performance and the dimensional stability of the fuel has also been developed within ABB. (author)

  3. PWR blowdown heat transfer separate-effects program: Thermal-Hydraulic Test Facility experimental data report for test 100

    International Nuclear Information System (INIS)

    White, M.D.; Hedrick, R.A.

    1977-01-01

    Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 100, which is part of the ORNL Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant accident in a PWR system. Test 100 was conducted to investigate the response of heater rod bundle 1 and instrumented spool pieces with flow homogenizing screens to a double-ended rupture with equal break areas at the test section inlet and outlet. The primary purpose of this report is to make the reduced instrument responses during test 100 available. The responses are presented in graphical form in engineering units and have been analyzed only to the extent necessary to assure reasonableness and consistency

  4. PWR Blowdown Heat Transfer Separate-Effects Program. Thermal-Hydraulic Test Facility experimental data report for test 166S

    International Nuclear Information System (INIS)

    Clemons, V.D.; White, M.D.; Hedrick, R.A.

    1978-01-01

    Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 166S, which is part of the ORNL Pressurized-Water Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant accident in a PWR system. Test 166S was conducted to obtain thermal-hydraulic and CHF information in THTF bundle 1 with an intact hot leg. The primary purpose of this report is to make the reduced instrument responses during tests 166S available. These are presented in graphical form in engineering units and have been analyzed only to the extent necessary to ensure reasonableness and consistency

  5. [Typical atrial flutter: Diagnosis and therapy].

    Science.gov (United States)

    Thomas, Dierk; Eckardt, Lars; Estner, Heidi L; Kuniss, Malte; Meyer, Christian; Neuberger, Hans-Ruprecht; Sommer, Philipp; Steven, Daniel; Voss, Frederik; Bonnemeier, Hendrik

    2016-03-01

    Typical, cavotricuspid-dependent atrial flutter is the most common atrial macroreentry tachycardia. The incidence of atrial flutter (typical and atypical forms) is age-dependent with 5/100,000 in patients less than 50 years and approximately 600/100,000 in subjects > 80 years of age. Concomitant heart failure or pulmonary disease further increases the risk of typical atrial flutter.Patients with atrial flutter may present with symptoms of palpitations, reduced exercise capacity, chest pain, or dyspnea. The risk of thromboembolism is probably similar to atrial fibrillation; therefore, the same antithrombotic prophylaxis is required in atrial flutter patients. Acutely symptomatic cases may be subjected to cardioversion or pharmacologic rate control to relieve symptoms. Catheter ablation of the cavotricuspid isthmus represents the primary choice in long-term therapy, associated with high procedural success (> 97 %) and low complication rates (0.5 %).This article represents the third part of a manuscript series designed to improve professional education in the field of cardiac electrophysiology. Mechanistic and clinical characteristics as well as management of isthmus-dependent atrial flutter are described in detail. Electrophysiological findings and catheter ablation of the arrhythmia are highlighted.

  6. Seismic qualification of PWR plant auxiliary feedwater systems

    International Nuclear Information System (INIS)

    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

  7. Development of Cost Estimation Methodology of Decommissioning for PWR

    International Nuclear Information System (INIS)

    Lee, Sang Il; Yoo, Yeon Jae; Lim, Yong Kyu; Chang, Hyeon Sik; Song, Geun Ho

    2013-01-01

    The permanent closure of nuclear power plant should be conducted with the strict laws and the profound planning including the cost and schedule estimation because the plant is very contaminated with the radioactivity. In Korea, there are two types of the nuclear power plant. One is the pressurized light water reactor (PWR) and the other is the pressurized heavy water reactor (PHWR) called as CANDU reactor. Also, the 50% of the operating nuclear power plant in Korea is the PWRs which were originally designed by CE (Combustion Engineering). There have been experiences about the decommissioning of Westinghouse type PWR, but are few experiences on that of CE type PWR. Therefore, the purpose of this paper is to develop the cost estimation methodology and evaluate technical level of decommissioning for the application to CE type PWR based on the system engineering technology. The aim of present study is to develop the cost estimation methodology of decommissioning for application to PWR. Through the study, the following conclusions are obtained: · Based on the system engineering, the decommissioning work can be classified as Set, Subset, Task, Subtask and Work cost units. · The Set and Task structure are grouped as 29 Sets and 15 Task s, respectively. · The final result shows the cost and project schedule for the project control and risk management. · The present results are preliminary and should be refined and improved based on the modeling and cost data reflecting available technology and current costs like labor and waste data

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

    International Nuclear Information System (INIS)

    Kim, Kyu-Tae

    2013-01-01

    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 −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

  9. Benefits of Low Boron Core Design Concept for PWR

    Energy Technology Data Exchange (ETDEWEB)

    Daing, Aung Tharn; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

    2009-10-15

    Nuclear design study was carried out to develop low boron core (LBC) based on one of current PWR concepts, OPR-1000. Most of design parameters were the same with those of Ulchin unit-5 except extensive utilization of burnable poison (BP) pins in order to compensate reactivity increase in LBC. For replacement of reduced soluble boron concentration, four different kinds of integral burnable absorbers (IBAs) such as gadolinia, integral fuel burnable absorber (IFBA), erbia and alumina boron carbide were considered in suppressing more excess reactivity. A parametric study was done to find the optimal core options from many design candidates for fuel assemblies and cores. Among them, the most feasible core design candidate was chosen in accordance with general design requirements. In this paper, the feasibility and design change benefits of the most favorable LBC design were investigated in more detail through the comparison of neutronic and thermal hydraulic design parameters of LBC with the reference plant (REF). As calculation tools, the HELIOS/MASTER code package and the MATRA code were utilized. The main purpose of research herein is to estimate feasibility and capability of LBC which was mainly designed to mitigate boron dilution accident (BDA), and for reduction of corrosion products. The LBC design concept using lower boron concentration with an elevated enrichment in {sup 10}B allows a reduction in the concentration of lithium in the primary coolant required to maintain the optimum coolant pH. All in all, LBC with operation at optimum pH is expected to achieve some benefits from radiation source reduction of reduced corrosion product, the limitation of the Axial Offset Anomaly (AOA) and fuel cladding corrosion. Additionally, several merits of LBC are closely related to fluid systems and system related aspects, reduced boron and lithium costs, equipment size reduction for boric acid systems, elimination of heat tracing, and more aggressive fuel design concepts.

  10. Benefits of Low Boron Core Design Concept for PWR

    International Nuclear Information System (INIS)

    Daing, Aung Tharn; Kim, Myung Hyun

    2009-01-01

    Nuclear design study was carried out to develop low boron core (LBC) based on one of current PWR concepts, OPR-1000. Most of design parameters were the same with those of Ulchin unit-5 except extensive utilization of burnable poison (BP) pins in order to compensate reactivity increase in LBC. For replacement of reduced soluble boron concentration, four different kinds of integral burnable absorbers (IBAs) such as gadolinia, integral fuel burnable absorber (IFBA), erbia and alumina boron carbide were considered in suppressing more excess reactivity. A parametric study was done to find the optimal core options from many design candidates for fuel assemblies and cores. Among them, the most feasible core design candidate was chosen in accordance with general design requirements. In this paper, the feasibility and design change benefits of the most favorable LBC design were investigated in more detail through the comparison of neutronic and thermal hydraulic design parameters of LBC with the reference plant (REF). As calculation tools, the HELIOS/MASTER code package and the MATRA code were utilized. The main purpose of research herein is to estimate feasibility and capability of LBC which was mainly designed to mitigate boron dilution accident (BDA), and for reduction of corrosion products. The LBC design concept using lower boron concentration with an elevated enrichment in 10 B allows a reduction in the concentration of lithium in the primary coolant required to maintain the optimum coolant pH. All in all, LBC with operation at optimum pH is expected to achieve some benefits from radiation source reduction of reduced corrosion product, the limitation of the Axial Offset Anomaly (AOA) and fuel cladding corrosion. Additionally, several merits of LBC are closely related to fluid systems and system related aspects, reduced boron and lithium costs, equipment size reduction for boric acid systems, elimination of heat tracing, and more aggressive fuel design concepts

  11. Status on development and verification of reactivity initiated accident analysis code for PWR (NODAL3)

    International Nuclear Information System (INIS)

    Peng Hong Liem; Surian Pinem; Tagor Malem Sembiring; Tran Hoai Nam

    2015-01-01

    A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the nodal few-group neutron diffusion theory in 3-dimensional Cartesian geometry for a typical pressurized water reactor (PWR) static and transient analyses, especially for reactivity initiated accidents (RIA). The spatial variables are treated by using a polynomial nodal method (PNM) while for the neutron dynamic solver the adiabatic and improved quasi-static methods are adopted. A simple single channel thermal-hydraulics module and its steam table is implemented into the code. Verification works on static and transient benchmarks are being conducting to assess the accuracy of the code. For the static benchmark verification, the IAEA-2D, IAEA-3D, BIBLIS and KOEBERG light water reactor (LWR) benchmark problems were selected, while for the transient benchmark verification, the OECD NEACRP 3-D LWR Core Transient Benchmark and NEA-NSC 3-D/1-D PWR Core Transient Benchmark (Uncontrolled Withdrawal of Control Rods at Zero Power). Excellent agreement of the NODAL3 results with the reference solutions and other validated nodal codes was confirmed. (author)

  12. A survey of available margin in a PWR RIA with statistical methods and 3D kinetics

    International Nuclear Information System (INIS)

    Riverola Gurruchaga, J.; Nunez Rodriguez, T.

    2010-01-01

    This paper investigates the recovery of margin in a PWR RIA simulation with 3D kinetics, due to statistical techniques. The chosen reference core is a typical 12 feet, 17*17 PWR, with very low leakage loading pattern strategy and gadolinium oxide as burnable poison. The PARCS calculated average nuclear power and nodal power are transferred to a hot spot model for a sequential calculation of fuel temperature and enthalpy responses allowing for independent hypothesis in both calculations. The hot spot analysis is done with a pellet type model with RELAP. The analysis is done at HZP and EOC, since this state is the most limiting one respect to the enthalpy rise criterion, compared to other burn-up condition or initial power cases. In this work, the enthalpy increase is estimated with several statistical methods of propagation of uncertainties: order statistics, parametric statistics, surface response and sensitivities. A discussion on the advantages and disadvantages of each method is also presented. This statistical analysis is also useful to confirm a previous classification of parameters and assumptions according to their importance for the simulation, and found to be consistent with the state of the art in the published literature. These parameters include ejected rod worth and ejection time, delayed neutron fraction and yields, nuclear power peaking factor, and Doppler. (authors)

  13. The role of phosphorus in the irradiation embrittlement of PWR pressure vessel steels

    International Nuclear Information System (INIS)

    Jones, R.B.; Buswell, J.T.

    1987-02-01

    An analysis has been performed of the influence of phosphorus on post-irradiation materials properties and microstructures determined on a variety of PWR steels and variants following exposure to MTR or reactor surveillance irradiations to doses not exceeding 7 x 10 19 n.cm -2 (E>1.0MeV) at 250-290 0 C. The irradiation-induced shifts in impact transition temperature, matrix hardening and the relative small angle neutron scattering response were found to rise most rapidly with increasing phosphorus when the copper content of the steel was 0.03 w/o. The sensitivity of the changes in mechanical properties to phosphorus content decreased as the copper content was increased. At copper levels typical of modern PWR steel manufacture (Cu 3 P) produced by the irradiation induced segregation of phosphorus to defect sinks and the depletion of phosphorus in solid solution as detected by high sensitivity electron microscopy and other analytical techniques. At higher levels of copper (approx. 0.3 w/o) the effect of phosphorus on properties was reduced by a factor of three due to the observed incorporation of phosphorus into the small copper precipitates formed during irradiation. Grain boundary embrittlement by phosphorus under irradiation is not thought to be important but further evidence concerning the post-irradiation fracture mode and the development of the deleterious influence of phosphorus with irradiation dose is required for a comprehensive understanding of its action. Some suggestions for future work are made. (author)

  14. Analysis of the return to power scenario following a LBLOCA in a PWR

    Energy Technology Data Exchange (ETDEWEB)

    Macian, R.; Tyler, T.N.; Mahaffy, J.H. [Pennsylvania State Univ., University Park, PA (United States)

    1995-09-01

    The risk of reactivity accidents has been considered an important safety issue since the beginning of the nuclear power industry. In particular, several events leading to such scenarios for PWR`s have been recognized and studied to assess the potential risk of fuel damage. The present paper analyzes one such event: the possible return to power during the reflooding phase following a LBLOCA. TRAC-PF1/MOD2 coupled with a three-dimensional neutronic model of the core based on the Nodal Expansion Method (NEM) was used to perform the analysis. The system computer model contains a detailed representation of a complete typical 4-loop PWR. Thus, the simulation can follow complex system interactions during reflooding, which may influence the neutronics feedback in the core. Analyses were made with core models bases on cross sections generated by LEOPARD. A standard and a potentially more limiting case, with increased pressurizer and accumulator inventories, were run. In both simulations, the reactor reaches a stable state after the reflooding is completed. The lower core region, filled with cold water, generates enough power to boil part of the incoming liquid, thus preventing the core average liquid fraction from reaching a value high enough to cause a return to power. At the same time, the mass flow rate through the core is adequate to maintain the rod temperature well below the fuel damage limit.

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

    International Nuclear Information System (INIS)

    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 ∞ 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

  16. A genetic algorithm applied to a PWR turbine extraction optimization to increase cycle efficiency

    International Nuclear Information System (INIS)

    Sacco, Wagner F.; Schirru, Roberto

    2002-01-01

    In nuclear power plants feedwater heaters are used to heat feedwater from its temperature leaving the condenser to final feedwater temperature using steam extracted from various stages of the turbines. The purpose of this process is to increase cycle efficiency. The determination of the optimal fraction of mass flow rate to be extracted from each stage of the turbines is a complex optimization problem. This kind of problem has been efficiently solved by means of evolutionary computation techniques, such as Genetic Algorithms (GAs). GAs, which are systems based upon principles from biological genetics, have been successfully applied to several combinatorial optimization problems in nuclear engineering, as the nuclear fuel reload optimization problem. We introduce the use of GAs in cycle efficiency optimization by finding an optimal combination of turbine extractions. In order to demonstrate the effectiveness of our approach, we have chosen a typical PWR as case study. The secondary side of the PWR was simulated using PEPSE, which is a modeling tool used to perform integrated heat balances for power plants. The results indicate that the GA is a quite promising tool for cycle efficiency optimization. (author)

  17. Optimization of refueling-shuffling scheme in PWR core by random search strategy

    International Nuclear Information System (INIS)

    Wu Yuan

    1991-11-01

    A random method for simulating optimization of refueling management in a pressurized water reactor (PWR) core is described. The main purpose of the optimization was to select the 'best' refueling arrangement scheme which would produce maximum economic benefits under certain imposed conditions. To fulfill this goal, an effective optimization strategy, two-stage random search method was developed. First, the search was made in a manner similar to the stratified sampling technique. A local optimum can be reached by comparison of the successive results. Then the other random experiences would be carried on between different strata to try to find the global optimum. In general, it can be used as a practical tool for conventional fuel management scheme. However, it can also be used in studies on optimization of Low-Leakage fuel management. Some calculations were done for a typical PWR core on a CYBER-180/830 computer. The results show that the method proposed can obtain satisfactory approach at reasonable low computational cost

  18. PTL: A Propositional Typicality Logic

    CSIR Research Space (South Africa)

    Booth, R

    2012-09-01

    Full Text Available consequence relations first studied by Lehmann and col- leagues in the 90?s play a central role in nonmonotonic reasoning [13, 14]. This has been the case due to at least three main reasons. Firstly, they are based on semantic constructions that are elegant...) j ; 6j : ^ j PTL: A Propositional Typicality Logic 3 The semantics of (propositional) rational consequence is in terms of ranked models. These are partially ordered structures in which the ordering is modular. Definition 1. Given a set S...

  19. PWR Facility Dose Modeling Using MCNP5 and the CADIS/ADVANTG Variance-Reduction Methodology

    Energy Technology Data Exchange (ETDEWEB)

    Blakeman, Edward D [ORNL; Peplow, Douglas E. [ORNL; Wagner, John C [ORNL; Murphy, Brian D [ORNL; Mueller, Don [ORNL

    2007-09-01

    The feasibility of modeling a pressurized-water-reactor (PWR) facility and calculating dose rates at all locations within the containment and adjoining structures using MCNP5 with mesh tallies is presented. Calculations of dose rates resulting from neutron and photon sources from the reactor (operating and shut down for various periods) and the spent fuel pool, as well as for the photon source from the primary coolant loop, were all of interest. Identification of the PWR facility, development of the MCNP-based model and automation of the run process, calculation of the various sources, and development of methods for visually examining mesh tally files and extracting dose rates were all a significant part of the project. Advanced variance reduction, which was required because of the size of the model and the large amount of shielding, was performed via the CADIS/ADVANTG approach. This methodology uses an automatically generated three-dimensional discrete ordinates model to calculate adjoint fluxes from which MCNP weight windows and source bias parameters are generated. Investigative calculations were performed using a simple block model and a simplified full-scale model of the PWR containment, in which the adjoint source was placed in various regions. In general, it was shown that placement of the adjoint source on the periphery of the model provided adequate results for regions reasonably close to the source (e.g., within the containment structure for the reactor source). A modification to the CADIS/ADVANTG methodology was also studied in which a global adjoint source is weighted by the reciprocal of the dose response calculated by an earlier forward discrete ordinates calculation. This method showed improved results over those using the standard CADIS/ADVANTG approach, and its further investigation is recommended for future efforts.

  20. PWR Facility Dose Modeling Using MCNP5 and the CADIS/ADVANTG Variance-Reduction Methodology

    International Nuclear Information System (INIS)

    Blakeman, Edward D.; Peplow, Douglas E.; Wagner, John C.; Murphy, Brian D.; Mueller, Don

    2007-01-01

    The feasibility of modeling a pressurized-water-reactor (PWR) facility and calculating dose rates at all locations within the containment and adjoining structures using MCNP5 with mesh tallies is presented. Calculations of dose rates resulting from neutron and photon sources from the reactor (operating and shut down for various periods) and the spent fuel pool, as well as for the photon source from the primary coolant loop, were all of interest. Identification of the PWR facility, development of the MCNP-based model and automation of the run process, calculation of the various sources, and development of methods for visually examining mesh tally files and extracting dose rates were all a significant part of the project. Advanced variance reduction, which was required because of the size of the model and the large amount of shielding, was performed via the CADIS/ADVANTG approach. This methodology uses an automatically generated three-dimensional discrete ordinates model to calculate adjoint fluxes from which MCNP weight windows and source bias parameters are generated. Investigative calculations were performed using a simple block model and a simplified full-scale model of the PWR containment, in which the adjoint source was placed in various regions. In general, it was shown that placement of the adjoint source on the periphery of the model provided adequate results for regions reasonably close to the source (e.g., within the containment structure for the reactor source). A modification to the CADIS/ADVANTG methodology was also studied in which a global adjoint source is weighted by the reciprocal of the dose response calculated by an earlier forward discrete ordinates calculation. This method showed improved results over those using the standard CADIS/ADVANTG approach, and its further investigation is recommended for future efforts

  1. Intergranular stress corrosion cracking of ion irradiated 304L stainless steel in PWR environment

    International Nuclear Information System (INIS)

    Gupta, Jyoti

    2016-01-01

    IASCC is irradiation - assisted enhancement of intergranular stress corrosion cracking susceptibility of austenitic stainless steel. It is a complex degrading phenomenon which can have a significant influence on maintenance time and cost of PWRs' core internals and hence, is an issue of concern. Recent studies have proposed using ion irradiation (to be specific, proton irradiation) as an alternative of neutron irradiation to improve the current understanding of the mechanism. The objective of this study was to investigate the cracking susceptibility of irradiated SA 304L and factors contributing to cracking, using two different ion irradiations; iron and proton irradiations. Both resulted in generation of point defects in the microstructure and thereby causing hardening of the SA 304L. Material (unirradiated and iron irradiated) showed no susceptibility to intergranular cracking on subjection to SSRT with a strain rate of 5 * 10 -8 s -1 up to 4 % plastic strain in inert environment. But, irradiation (iron and proton) was found to increase intergranular cracking severity of material on subjection to SSRT in simulated PWR primary water environment at 340 C. Correlation between the cracking susceptibility and degree of localization was studied. Impact of iron irradiation on bulk oxidation of SA 304L was studied as well by conducting an oxidation test for 360 h in simulated PWR environment at 340 C. The findings of this study indicate that the intergranular cracking of 304L stainless steel in PWR environment can be studied using Fe irradiation despite its small penetration depth in material. Furthermore, it has been shown that the cracking was similar in both iron and proton irradiated samples despite different degrees of localization. Lastly, on establishing iron irradiation as a successful tool, it was used to study the impact of surface finish and strain paths on intergranular cracking susceptibility of the material. (author) [fr

  2. 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.

  3. The pseudo-harmonics method: an application involving perturbations caused by control rod insertion in PWR reactors

    International Nuclear Information System (INIS)

    Claro, L.H.; Alvim, A.C.M.; Thome, Z.D.

    1988-08-01

    The objective of this work is to stydy the effect of intense perturbations, such as control rod insertion in the core of PWR reactors, through a perturbation approach consisting of a modified version of the pseudo-harmonics method. A typical one-dimensional PWR reactor model was used as a reference state, from which two perturbations were imposed, simulation gray and black control rod insertion. In the first case, eigenvalue convergence was achieved with the eighth order of approximation approximation and perturbed fluxes and eigenvalue estimates agreed very well with direct calculation results. The second case tested represents a very intense localized perturbation. Oscillation in keff were observed er of approximation increased and the method failed to converge. Results obtained indicate that the pseudo-harmonics method can be used to compute 2 group fluxes and fundamental eigenvalue of perturbated states resulting from gray control rod insertion in PWR reactors. The method is limited, however, by perturbation intensity, as other perturbation methods are. (author) [pt

  4. Simulation of a Nuclear Steam Supply System (NSSS) of a PWR nuclear power plant. Simulacao do sistema nuclear de geracao de vapor de uma central PWR

    Energy Technology Data Exchange (ETDEWEB)

    Reis Martins Junior, L.L. dos.

    1980-01-01

    The following work intends to perform the digital simulation, of the Nuclear Steam Supply System (NSSS) of a PWR nuclear power plant for control systems design and analysis purposes. There are mathematical models for the reactor, the steam generator, the pressurizer and for transport lags of the coolant in the primary circuit. Nevertheless no one control system has been considered to permit any user the inclusion in the more convenient way of the desired control systems' models. The characteristics of the system in consideration are fundamentally equal to the ones of Almirante Alvaro Alberto Nuclear Power Plant, Unit I (Angra I) obtained in the Final Safety Analysis Report at Comissao Nacional de Energia Nuclear. (author).

  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.; 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)

  6. Load-following operation of PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Jong Hwa; Oh, Soo Yul; Koo, Yang Hyun; Lee, Jae Han [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-12-01

    The load-following operation of nuclear power plants will become inevitable due to the increased nuclear share in the total electricity generation. As a groundwork for the load-following capability of the Korean next generation PWRs, the state-of-the-art has been reviewed. The core control principles and methods are the main subject in this review as well as the impact of load-following operations on the fuel performance and on the mechanical integrity of components. To begin with, it was described what the load-following operation is and in what view point the technology should be reviewed. Afterwards the load-following method, performance and problems in domestic 900 MWe class PWRs were discussed, and domestic R and D works were summarized. Foreign technologies were also reviewed. They include Mode G and Mode X of Foratom, D and L bank method of KWU, the method using PSCEA of ABB-CE, and MSHIM of Westinghouse. The load-following related special features of Foratom`s N4 plant, KWU`s plants, ABB-CE`s Systems 80+, and Westinghouse`s AP600 were described in each technology review. The review concluded that the capability of N4 plant with Mode X is the best and the methods in System, 80+ and AP600 would require verifications for the continued and usual load-following operation. It was recommended that the load-following operation experiences in domestic PWRs under operation be required to settle down the capability for the future. In addition, a more enhanced technology is required for the Korean next generation PWR regardless what the reference plant concept is. 30 figs., 19 tabs., 75 refs. (Author).

  7. Basic information about development and construction of a PWR

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1977-01-01

    1.0) Plant layout of a PWR; 2.0) principle design of a PWR and the reactor coolant system; 3.0) reactor auxiliary and ancillary systems; 3.1) volume control system; 3.2) boric acid control and chemical feeding system; 3.3) coolant purification and degassing system; 3.4) coolant storage and treatment system; 3.5) nuclear component cooling system; 3.6) liquid waste processing system; 3.7) gaseous waste processing system; 4.0) residual heat removal system; 5.0) emergency feedwater system; 6.0) containment design; 7.0) fuel handling, storage and transport system in a PWR. (orig.) [de

  8. Improved emergency elevated air release for simplified PWR

    International Nuclear Information System (INIS)

    Naitoh, T.; Bruce, R.A.; Hirota, K.; Tajiri, Y.

    1992-01-01

    In developing the application of the simplified PWR in Japan, one of the most important areas is to limit post-accident site boundary whole body dose. In addressing this, the concept of Emergency Passive Air Filtration System (EPAFS) and it's feasibility is developed. The efficiency of charcoal filtering and the atmospheric diffusion effect of an elevated air release are important for dose reduction. The performance of these functions was evaluated by confirmatory testing. The test results confirmed a 99 percent efficiency of charcoal filter and an atmospheric diffusion effect higher than that of a conventional plant. The Emergency Passive Air Filtration System (EPAFS) and the atmospheric diffusion effect of elevated air release contribute to making the calculated post-accident site boundary whole body dose of simplified PWR as low as that of the conventional Japanese PWR plant. (author)

  9. Swing-Down of 21-PWR Waste Package

    International Nuclear Information System (INIS)

    A.K. Scheider

    2001-01-01

    The objective of this calculation is to determine the structural response of the waste package (WP) swinging down from a horizontally suspended 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. 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. 13). AP-3.12Q, ''Calculations'' (Ref. 18) is used to perform the calculation and develop the document. The information provided by the sketches attached to this calculation is that of the potential design of the type of 21-PWR WP design considered in this calculation and provides the potential dimensions and materials for the 21-PWR WP design

  10. Pressurizer and steam-generator behavior under PWR transient conditions

    International Nuclear Information System (INIS)

    Wahba, A.B.; Berta, V.T.; Pointner, W.

    1983-01-01

    Experiments have been conducted in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR), at the Idaho National Engineering Laboratory, in which transient phenomena arising from accident events with and without reactor scram were studied. The main purpose of the LOFT facility is to provide data for the development of computer codes for PWR transient analyses. Significant thermal-hydraulic differences have been observed between the measured and calculated results for those transients in which the pressurizer and steam generator strongly influence the dominant transient phenomena. Pressurizer and steam generator phenomena that occurred during four specific PWR transients in the LOFT facility are discussed. Two transients were accompanied by pressurizer inflow and a reduction of the heat transfer in the steam generator to a very small value. The other two transients were accompanied by pressurizer outflow while the steam generator behavior was controlled

  11. Fabrication of PWR fuel assembly and CANDU fuel bundle

    International Nuclear Information System (INIS)

    Lee, G.S.; Suh, K.S.; Chang, H.I.; Chung, S.H.

    1980-01-01

    For the project of localization of nuclear fuel fabrication, the R and D to establish the fabrication technology of CANDU fuel bundle as well as PWR fuel assembly was carried out. The suitable boss height and the prober Beryllium coating thickness to get good brazing condition of appendage were studied in the fabrication process of CANDU fuel rod. Basic Studies on CANLUB coating method also were performed. Problems in each fabrication process step and process flow between steps were reviewed and modified. The welding conditions for top and bottom nozzles, guide tube, seal and thimble screw pin were established in the fabrication processes of PWR fuel assembly. Additionally, some researches for a part of PWR grid brazing problems are also carried out

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

    International Nuclear Information System (INIS)

    Tsuchiya, A.; Ito, K.; Yokoyama, M.

    2001-01-01

    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)

  13. PWR fuel performance and future trend in Japan

    International Nuclear Information System (INIS)

    Kondo, Y.

    1987-01-01

    Since the first PWR power plant Mihama Unit 1 initiated its commercial operation in 1970, Japanese utilities and manufacturers have expended much of their resources and efforts to improve PWR technology. The results are already seen in significantly improved performance of 16 PWR plants now in operation. Mitsubishi Heavy Industries Ltd. (MHI) has been supplying them with nuclear fuel assemblies, which are over 5700. As the reliability of the current design fuel has been achieved, the direction of R and D on nuclear fuel has changed to make nuclear power more competitive to the other power generation methods. The most important R and D targets are the burnup extension, Gd contained fuel, Pu utilizatoin and the load follow capacility. (author)

  14. Radionuclide compositions of spent fuel and high level waste for the uranium and plutonium fuelled PWR

    International Nuclear Information System (INIS)

    Fairclough, M.P.; Tymons, B.J.

    1985-06-01

    The activities of a selection of radionuclides are presented for three types of reactor fuel of interest in radioactive waste management. The fuel types are for a uranium 'burning' PWR, a plutonium 'burning' PWR using plutonium recycled from spent uranium fuel and a plutonium 'burning' PWR using plutonium which has undergone multiple recycle. (author)

  15. Velocity of crack growing of Inconel-600, sensitized, contaminated with sulphur in PWR type reactors

    International Nuclear Information System (INIS)

    Castano, M. L.; Blazquez, F.; Gomez Briceno, D.; Lagares, A.

    1998-01-01

    The origin of the vessel head penetration cracking of Jose Cabrera NPP has been attributed to an IGA/SCC process in a highly sensitized Alloy 600 assisted by sulphur species, as both acid sulphates and reduced species originated by the thermal breakdown of the cationic resins present in the primary coolant. The thermal degradation of the cationic resins leads sulphonic acid group scission and sulphates. Under the operating conditions the reduction of sulphates to sulphides is produced. The sulphides formed from the reduction of sulphate can precipitate with metallic cations and be incorporated into the oxide layers of the materials, preferably into nickel alloys. Others components at Jose Cabrera NPP are fabricated from sensitized alloy 600, as bottom vessel penetrations. In order to determine the influence of sulphur incorporated to the oxide layers of bottom vessel penetration alloy 600, an experimental work has been performed to obtained crack growth rate data under PWR primary conditions on sensitized alloy 600. (Author) 5 refs

  16. 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)

  17. The traveller: a new look for PWR fresh fuel packages

    International Nuclear Information System (INIS)

    Bayley, B.; Stilwell, W.E.; Kent, N.A.

    2004-01-01

    The Traveller PWR fresh fuel shipping package represents a radical departure from conventional PWR fuel package designs. This paper follows the development effort from the establishment of goals and objectives, to intermediate testing and analysis, to final testing and licensing. The discussion starts with concept origination and covers the myriad iterations that followed until arriving at a design that would meet the demanding licensing requirements, last for 30 years, and would be easy to load and unload fuel, easy to handle, inexpensive to manufacture and transport, and simple and inexpensive to maintain

  18. The study on radioactivity reduction of spent PWR cladding hull

    International Nuclear Information System (INIS)

    Jung, I. H.; Kim, J. H.; Park, C. J.; Jung, Y. H.; Song, K. C.; Lee, J. W.; Park, J. J.; Yang, M. S.

    2003-01-01

    Hull arising from the spent PWR fuel elements is classified as a high-level radioactive waste. This report describes the radio-chemical characteristics of the hull-from PWR spent fuel of 32,000MWd/tU burn-up and 15 years cooling, discharged from Gori Unit I cycled 4-7-by examination and literature survey. On the basis of the results, a method of degradation to middle and low-level radioactive waste was proposed by dry process such as laser or plasma technique with removing the nuclides deposited on the surface of the hull

  19. BEACON TSM application system to the operation of PWR reactors

    International Nuclear Information System (INIS)

    Lozano, J. A.; Mildrum, C.; Serrano, J. F.

    2012-01-01

    BEACON-TSM is an advanced core monitoring system for PWR reactor cores, and also offers the possibility to perform a wide range of predictive calculation in support of reactor operation. BEACON-TSM is presently installed and licensed in the 5 Spanish PWR reactors of standard Westinghouse design. the purpose of this paper is to describe the features of this software system and to show the advantages obtainable by a nuclear power plant from its use. To illustrate the capabilities and benefits of BEACON-TSM two real case reactor operating situations are presented. (Author)

  20. Steam Generator Owners Group PWR secondary water chemistry guidelines

    International Nuclear Information System (INIS)

    Welty, C.S. Jr.; Green, S.J.

    1985-01-01

    In 1981 the Steam Generator Owners Group (SGOG), a group of domestic and foreign pressurized water reactor (PWR) owners, developed and issued the PWR secondary water chemistry guidelines. The guidelines were prepared in response to the growing recognition that a majority of the problems causing reduced steam generator reliability (e.g., denting, wasteage, pitting, etc.) were related to secondary (steam) side water purity. The guidelines were subsequently issued as an Electric Power Research Institute (EPRI) report. In 1984 they were revised to reflect industry experience in adopting the original issuance and to incorporate new information on causes of corrosion damage. The guidelines have been endorsed and their adoption recommended by the SGOG

  1. Sensitivity of risk parameters to human errors for a PWR

    International Nuclear Information System (INIS)

    Samanta, P.; Hall, R.E.; Kerr, W.

    1980-01-01

    Sensitivities of the risk parameters, emergency safety system unavailabilities, accident sequence probabilities, release category probabilities and core melt probability were investigated for changes in the human error rates within the general methodological framework of the Reactor Safety Study for a Pressurized Water Reactor (PWR). Impact of individual human errors were assessed both in terms of their structural importance to core melt and reliability importance on core melt probability. The Human Error Sensitivity Assessment of a PWR (HESAP) computer code was written for the purpose of this study

  2. The latest full-scale PWR simulator in Japan

    International Nuclear Information System (INIS)

    Nishimuru, Y.; Tagi, H.; Nakabayashi, T.

    2004-01-01

    The latest MHI Full-scale Simulator has an excellent system configuration, in both flexibility and extendability, and has highly sophisticated performance in PWR simulation by the adoption of CANAC-II and PRETTY codes. It also has an instructive character to display the plant's internal status, such as RCS condition, through animation. Further, the simulation has been verified to meet a functional examination at model plant, and with a scale model test result in a two-phase flow event, after evaluation for its accuracy. Thus, the Simulator can be devoted to a sophisticated and broad training course on PWR operation. (author)

  3. Advanced ion exchange resins for PWR condensate polishing

    International Nuclear Information System (INIS)

    Hoffman, B.; Tsuzuki, S.

    2002-01-01

    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)

  4. Two optimal control methods for PWR core control

    International Nuclear Information System (INIS)

    Karppinen, J.; Blomsnes, B.; Versluis, R.M.

    1976-01-01

    The Multistage Mathematical Programming (MMP) and State Variable Feedback (SVF) methods for PWR core control are presented in this paper. The MMP method is primarily intended for optimization of the core behaviour with respect to xenon induced power distribution effects in load cycle operation. The SVF method is most suited for xenon oscillation damping in situations where the core load is unpredictable or expected to stay constant. Results from simulation studies in which the two methods have been applied for control of simple PWR core models are presented. (orig./RW) [de

  5. Irradiation behavior of German PWR RPV steels under operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    May, J.; Hein, H. [AREVA NP Gmbh (Germany); Ganswind, J. [VGB PowerTech e.V. (Germany); Widera, M. [RWE Power AG (Germany)

    2011-07-01

    In 2007, the last standard surveillance capsule of the original RPV (Reactor Pressure Vessel) surveillance programs of the 11 currently operating German PWR has been evaluated. With it the standard irradiation surveillance programs of these plants was completed. In the present paper, irradiation data of these surveillance programs will be presented and a final assessment of the irradiation behavior of the German PWR RPV steels with respect to current standards KTA 3203 and Reg. Guide 1.99 Rev. 2 will be given. Data from two units which are currently under decommissioning will also be included, so that data from all 13 German PWR manufactured by the former Siemens/KWU company (now AREVA NP GmbH) are shown. It will be shown that all surveillance data within the approved area of chemical composition verify the limit curve RT(limit) of the KTA 3203, which is the relevant safety standard for these plants. An analysis of the data shows, that the prediction formulas of Reg. Guide 1.99 Rev. 2 Pos. 1 or from the TTS model tend to overestimate the irradiation behavior of the German PWR RPV steels. Possible reasons for this behavior are discussed. Additionally, the data will be compared to data from the research project CARISMA to demonstrate that these data are representative for the irradiation behavior of the German PWR RPV steels. Since the data of these research projects cover a larger neutron fluence range than the original surveillance data, they offer a future outlook into the irradiation behavior of the German PWR RPV steels under long term conditions. In general, as a consequence of the relatively large and beneficial water gap between core and RPV, especially in all Siemens/KWU 4-loop PWR, the EOL neutron fluence and therefore the irradiation induced changes in mechanical properties of the German PWR RPV materials are rather low. Moreover the irradiation data indicate that the optimized RPV materials specifications that have been applied in particular for the

  6. A Multi-Physics PWR Model for the Load Following

    OpenAIRE

    Muniglia , Mathieu; Do , Jean-Michel; Jean-Charles , Le Pallec; Grard , Hubert; Verel , Sébastien; David , S.

    2016-01-01

    International audience; In this paper, a new model of a Pressurized Water Reactor (PWR) is described. This model includes the description of the core as well as a simplified secondary loop: the goal is to reproduce a load-following type transient, where the output power of the plant is controlled by the electric grid. Consequently, the control systems are also modeled, as the control rods or the soluble boron. The reference power plant is a 1300MW electrical PWR, managed with the french G mode.

  7. Evaluation of Computational Fluids Dynamics (CFD) code Open FOAM in the study of the pressurized thermal stress of PWR reactors. Comparison with the commercial code Ansys-CFX

    International Nuclear Information System (INIS)

    Martinez, M.; Barrachina, T.; Miro, R.; Verdu Martin, G.; Chiva, S.

    2012-01-01

    In this work is proposed to evaluate the potential of the OpenFOAM code for the simulation of typical fluid flows in reactors PWR, in particular for the study of pressurized thermal stress. Test T1-1 has been simulated , within the OECD ROSA project, with the objective of evaluating the performance of the code OpenFOAM and models of turbulence that has implemented to capture the effect of the thrust forces in the case study.

  8. A Calculation of the radioactivity induced in PWR cluster control rods with the origin and casmo codes

    International Nuclear Information System (INIS)

    Ekberg, K.

    1980-03-01

    The radioactivity induced in PWR cluster control rods during reactor operation has been calculated using the computer programme ORIGEN. Neutron fluxes and spectrum conditions as well as the strongly shielded cross sections for the absorber materials Ag, In and Cd have been obtained by running the cell and assembly code CASMO for a couple of typical cases. The results show that Ag-110m, Fe-55 and Co-60 give the largest activity contributions in the interval 1-10 years after the end of irradiation, and Ni-63 and Cd-113m in a longer time perspective. (author)

  9. Assessment of some interfacial shear correlations in a model of ECC bypass flow in PWR reactor downcomer

    International Nuclear Information System (INIS)

    Popov, N.K.; Rohatgi, U.S.

    1987-01-01

    The bypass/refill process in the PWR reactor downcomer, following a large rupture of a cold leg coolant supply pipe, is a complicated thermo-hydraulic two-phase flow phenomenon. Mathematical modeling of such phenomena is always accompanied with a difficult task of selection of suitable constitutive correlations. In a typically hydrodynamic phenomenon, like ECC refill process of the reactor lower plenum is considered, the phasic interfacial friction is the most influential constitutive correlation. Therefore, assessment of the well-known widely-used interfacial friction constitutive correlations in the model of ECC bypass/refill process, is the subject of this paper

  10. Stress corrosion cracking in the vessel closure head penetrations of French PWR's

    International Nuclear Information System (INIS)

    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

  11. Ventilation and air-conditioning system for PWR nuclear power plant

    International Nuclear Information System (INIS)

    Ohmoto, Kenji

    1987-01-01

    This report outlines the ventilation and air conditioning facilities for PWR nuclear power plant as well as design re-evaluation and optimization of ventilation and air-conditioning. The primary PWR installations are generally housed in the nuclear reactor building, auxiliary buildings and control building, which are equipped with their own ventilation and air-conditioning systems to serve for their specific purposes. A ventilation/air-conditioning system should be able to work effectively not only for maintaining the ordinary reactor operation but also for controlling the environmental temperature in the event of an accident. Designing of a ventilation/air-conditioning system relied on empirical data in the past, but currently it is performed based on information obtained from various analyses to optimize the system configuration and ventilation capacity. Design re-evaluation of ventilation/air-conditioning systems are conducted widely in various areas, aiming at the integration of safety systems, optimum combination of air-cooling and water-cooling systems, and optimization of the ventilation rate for controlling the concentrations of radioactive substances in the atmosphere in the facilities. It is pointed out that performance evaluation of ventilation/air-conditioning systems, which has been conducted rather macroscopically, should be carried out more in detal in the future to determine optimum air streams and temperature distribution. (Nogami, K.)

  12. A comparison of Zircaloy oxide thicknesses on Millstone-3 and North Anna-1 PWR fuel cladding

    International Nuclear Information System (INIS)

    Polley, M.V.; Evans, H.E.

    1993-08-01

    High concentrations of lithium in the coolant may enhance the corrosion rate of Zircaloy fuel cladding. In the present work, oxide thicknesses on fuel cladding from the Millstone 3 PWR were compared with those from the North Anna 1 PWR. The intention was to identify whether the higher lithium levels (up to 3.5 ppM) in the Millstone 3 primary coolant during cycles 2 and 3 led to significantly greater oxidation rates than in North Anna 1 which operated generally with lithium levels lower than 2.2 ppM. The comparisons were made by comparing the measurements with code predictions of Zircaloy oxidation in order to factor out the effect of operational variables on the oxide thicknesses achieved. Overall, Millstone 3 oxide thicknesses were found to be approximately 14% greater than North Anna 1 values. However, approximately 29% lower oxide thicknesses were found on reload Millstone 3 rods exposed to one cycle of elevated lithium chemistry than on Millstone 3 initial fuel exposed to one cycle of normal lithium chemistry during cycle 1. Furthermore, oxide thicknesses on Millstone 3 rods exposed to two cycles of elevated lithium chemistry were approximately 36% lower than on Millstone 3 rods exposed to one cycle of normal lithium chemistry plus one cycle of elevated lithium chemistry. Therefore, it cannot be concluded that elevated lithium operation in Millstone 3 led to enhanced Zircaloy fuel clad corrosion

  13. Failure of PWR-RHRS under cold shutdown conditions: Experimental results from the PKL test facility

    International Nuclear Information System (INIS)

    Mandl, R.M.; Umminger, K.J.; Logt, J.V.D.

    1991-01-01

    The Residual Heat Removal System (RHRS) of a PWR is designed to transfer thermal energy from the core after plant shutdown and maintain the plant in cold shutdown or refuelling conditions for extended periods of time. Initial reactor cooling after shutdown is achieved by dissipating heat through the steam generators (SGs) and discharging steam to the condenser by means of the Turbine Bypass System (TBS). When the reactor coolant temperature has dropped to about 160C and pressure has been reduced to 30 bar the RHRS is placed into operation. it reduces the coolant temperature to 50C within 20 hours after shutdown. The time margin for establishing alternate methods of heat removal following a failure of the RHRS depends on the Reactor Coolant System (RCS) temperature, the decay heat rate and the amount of RCS inventory. During some shutdown operations the RCS may be partially drained (e. g. to perform SG inspections). Decreased primary system inventory can significantly reduce the time available to recover the RHRS's function prior to bulk boiling and possible core uncovery. In the PKL test facility, which simulates a 1,300 MWe 4-loop PWR on a scale 1:145, a failure of RHRS under cold shutdown conditions was performed. This presentation gives a brief description of the test facility followed by the test objectives and results of this experiment

  14. Simulation of small break loss of coolant accident in pressurized water reactor (PWR)

    International Nuclear Information System (INIS)

    Abass, N. M. N.

    2012-02-01

    A major safety concern in pressurized-water-reactor (PWR) design is the loss-of-coolant accident (LOCA),in which a break in the primary coolant circuit leads to depressurization, boiling of the coolant, consequent reduced cooling of the reactor core, and , unless remedial measures are taken, overheating of the fuel rods. This concern has led to the development of several simulators for safety analysis. This study demonstrates how the passive and active safety systems in conventional and advanced PWR behave during the small break loss of Coolant Accident (SBLOCA). The consequences of SBOLOCA have been simulated using IAEA Generic pressurized Water Reactor Simulator (GPWRS) and personal Computer Transient analyzer (PCTRAN) . The results were presented and discussed. The study has confirmed the major safety advantage of passive plants versus conventional PWRs is that the passive safety systems provide long-term core cooling and decay heat removal without the need for operator actions and without reliance on active safety-related system. (Author)

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

    International Nuclear Information System (INIS)

    Febrianto

    2006-01-01

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

  16. Fuel assembly loads during a hypothetical blowdown event in a PWR

    International Nuclear Information System (INIS)

    Stabel, J.; Bosanyi, B.; Kim, J.D.

    1991-01-01

    As a consequence of a hypothetical sudden break of the main coolant pipe of a PWR, RPV-internals and fuel assemblies (FA's) are undergoing horizontal and vertical motions. FA's may impact against each other, against core shroud or against lower core support. The corresponding impact loads must be absorbed by the FA spacer grids and guide thimbles. In this paper FA-loads are calculated with and without consideration of Fluid-Structure-Interaction (FSI) effects for assumed different break sizes of the main coolant pipe. The analysis has been performed for a hypothetical cold leg break of a typical SIEMENS-4 loop plant. For this purpose the codes DAPSY/DAISY (GRS, Germany) were coupled with the structural code KWUSTOSS (SIEMENS). It is shown that the FA loads obtained in calculations with consideration of FSI effects are by a factor of 2-4 lower than those obtained in the corresponding calculations without consideration of FSI. (author)

  17. Theoretical-experimental modelling of the momentum equation for PWR reactor steam generators

    International Nuclear Information System (INIS)

    Rodrigues, L.A.H.

    1994-01-01

    A mathematical model in steady-state conditions of the momentum equation at the secondary side of a vertical U-tube steam generator with recirculation is presented. The U-tube test section was the 150 bar - Circuito Termoidraulico Experimental - CTE-150. This facility is a Experimental Thermal-hydraulic Circuit and operates at the same conditions (pressure and temperature) of a typical PWR reactor. A comparison between the Homogeneous and Separate Flow models was done. those models were verified and compared with experimental data for several operational conditions. The results show that the model fits very well the experimental data and seems to be appropriate to study water recirculation of a steam generator secondary side. (author)

  18. Simulating the steam generator and the pressurizer of a PWR nuclear power plant

    International Nuclear Information System (INIS)

    De Greef, J.F.

    1985-01-01

    In a PWR nuclear power plant, considered as a power generating device, the steam generator as a subset plays an important role in the generation process, whereas the pressurizer rather acts as a control device for security purposes. Nevertheless, from a thermodynamical point of view, the two subsets behave basically in the same way, so that a common set of basic equations may be suggested to develop for each the proper mathematical simulation model. In this paper the generation of this common set of basic equations is described, from which a specific model for each device is derived. A numerical illustration of the behaviour of the two devices for typical inputs to the derived simulation model is pictured. (author)

  19. Turbulent heat transfer in a coolant channel of a pressurized water reactor (PWR) core

    International Nuclear Information System (INIS)

    Kumar, Sanjeev; Saha, Arun K.; Munshi, Prabhat

    2016-01-01

    Exact predictions in nuclear reactors are more crucial, because of the safety aspects. It necessitates the appropriate modeling of heat transfer phenomena in the reactors core. A two-dimensional thermal-hydraulics model is used to study the detailed analysis of the coolant region of a fuel pin. Governing equations are solved using Marker and Cell (MAC) method. Standard wall functions k-ε turbulence model is incorporated to consider the turbulent behaviour of the flow field. Validation of the code and a few results for a typical PWR running at normal operating conditions reported earlier. There were some discrepancies in the old calculations. These discrepancies have been resolved and updated results are presented in this work. 2D thermal-hydraulics model results have been compared with the 1D thermal-hydraulics model results and conclusions have been drawn. (author)

  20. Advanced PWR Core Design with Siemens High-Plutonium-Content MOX Fuel Assemblies

    International Nuclear Information System (INIS)

    Dieter Porsch; Gerhard Schlosser; Hans-Dieter Berger

    2000-01-01

    The Siemens experience with plutonium recycling dates back to the late 1960s. Over the years, extensive research and development programs were performed for the qualification of mixed-oxide (MOX) technology and design methods. Today's typical reload enrichments for uranium and MOX fuel assemblies and modern core designs have become more demanding with respect to accuracy and reliability of design codes. This paper presents the status of plutonium recycling in operating high-burnup pressurized water reactor (PWR) cores. Based on actual examples, it describes the validation status of the design methods and stresses current and future needs for fuel assembly and core design including those related to the disposition of weapons-grade plutonium