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Sample records for 5x5 pwr rod

  1. Numerical evaluation of flow through a 5X5 PWR rod bundle: effect of the vane arrangement in a spacer grid

    Spacer grids along the fuel assembly of Pressurized Water Reactors (PWR) maintain rod bundles arranged in a regular square configuration. The mixing vanes present in the spacer grids promote cross and swirl flow between and within the subchannels, enhancing the heat transfer performance in the grid vicinity, but also causing an adverse increase of the pressure drop in the rod bundle due the constriction on the coolant flow area. Therefore, the thermal hydraulic design of the grid must allow for both low pressure loss and high coolant mixing, which means it is important to optimize the design of the grid in relation to the mixing vane. More recently, Computational Fluid Dynamics (CFD) using three dimensional Reynolds Averaged Navier Stokes (RANS) analysis has been used efficiently as an auxiliary tool in the development of spacer grids. The influence of some geometric characteristics of spacer grids on the flow through a rod bundle have been numerically evaluated and are still a subject of discussion. This work analyses the influence of the vanes arrangement in the spacer grid on the flow through a PWR 5 x 5 rod bundle segment. The Numerical simulations were performed with the commercial code CFX 11.0. To make the simulation possible with a limited computational capacity and acceptable mesh refinement, the computational domain was divided in 7 subdomains. The subdomains were simulated sequentially applying the outlet results of a previous subdomain as inlet condition for the next. In this study the k- turbulence model with scalable wall function was used. Five different vane arrangements were simulated at reactor level power and flow characteristics. The same grid and vane geometry were used in all simulations. The results of this study were divided in two parts. In the first part the presence of peripheral vanes on 5 x 5 rod bundle spacer grid tests were evaluated. The results showed that peripheral vanes should be avoided in experiments and simulations in order to

  2. Evaluation of a numeric procedure for flow simulation of a 5X5 PWR rod bundle with a mixing vane spacer

    The fuel assemblies of the Pressurized Water Reactors (PWR) are constituted of rod bundles arranged in a regular square configuration by spacer grids placed along its length. The presence of the spacer grids promote two antagonist effects on the core: a desirable increase of the local heat transfer downstream the grids and an adverse increase of the pressure drop due the constriction on the coolant flow area. Most spacer grids are designed with mixing vanes which cause a cross and swirl flow between and within the subchannels, enhancing even more the heat transfer performance in the grid vicinity. The improvement of the heat transfer increases the departure from the nucleate boiling ratio, allowing higher operating power in the reactor. Due to these important thermal and fluid dynamic features, experimental and theoretical investigations have been carried out in the past years for the development of spacer grid design. More recently, the Computational Fluid Dynamics (CFD) using three dimensional Reynolds Averaged Navier Stokes (RANS) analysis has been used efficiently for this purpose. Many computational works have been performed, but the appropriate numerical procedure for the flow in rod bundle simulations is not yet a consensus. This work presents results of flow simulations performed with the commercial code CFX 11.0 in a PWR 5x5 rod bundle segment with a split vane spacer grid. The geometrical configuration and flow conditions used in the experimental studies performed by Karoutas et al. were assumed in the simulations. To make the simulation possible with a limited computational capacity and acceptable mesh refinement, the computational domain was divided in 7 subdomains. The subdomains were simulated sequentially applying the outlet results of a previous subdomain as inlet condition for the next. In this study the κ-ε turbulence model was used. The simulations were also compared with those performed by Karoutas et al. in half a subchannel and In et al. in

  3. Reflood Phenomena in a 5 x 5 Ballooned Rod Bundle

    Kim, Byoung Jae; Kim, Jong Rok; Kim, Kihwan; Moon, S. K. [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    Various experimental programs were carried out for the coolability of an assembly containing a partial blockage in a group of ballooned fuel rods under LOCA conditions. A review on these experimental programs is well documented in. One key distinguished feature of KAERI research activities is the consideration of local power increase owing to fuel relocation, whereas the past experimental program did not consider the effect of fuel relocation. The purpose of this study is to investigate the reflood phenomena in the partial blocked 5 x 5 rod bundle. A series of the forced reflood tests were performed with/without consideration of local power increase by fuel relocation. The experimental data were evaluated with numerical predictions using MARS code. The flow blockage alone has little effect on the peak wall temperature. However, the local power increase by fuel relocation affects considerably the peak wall temperature and the time period during which high wall temperatures continue.

  4. Reflood Phenomena in a 5 x 5 Ballooned Rod Bundle

    Various experimental programs were carried out for the coolability of an assembly containing a partial blockage in a group of ballooned fuel rods under LOCA conditions. A review on these experimental programs is well documented in. One key distinguished feature of KAERI research activities is the consideration of local power increase owing to fuel relocation, whereas the past experimental program did not consider the effect of fuel relocation. The purpose of this study is to investigate the reflood phenomena in the partial blocked 5 x 5 rod bundle. A series of the forced reflood tests were performed with/without consideration of local power increase by fuel relocation. The experimental data were evaluated with numerical predictions using MARS code. The flow blockage alone has little effect on the peak wall temperature. However, the local power increase by fuel relocation affects considerably the peak wall temperature and the time period during which high wall temperatures continue

  5. CFD Verification of 5x5 Rod Bundle with Mixing Vane Spacer Grids

    Results of the CHF test are used for determining the CHF correlation, which is used to evaluate the thermal margin in the reactor core. Computational fluid dynamics (CFD) has been used to save the time and cost for experimental tests, components design and complicated phenomena in all industries including the reactor coolant system. L. D. Smith et al. applied the CFD methodology in a 5x5 rod bundle with the mixing vane spacer grid using the renormalization group (RNG) k-epsilon model. This CFD model agreed reasonably well with the test data. M. E. Conner et al. conducted experiments to validate the CFD methodology for the single-phase flow conditions in PWR fuel assemblies. In this validation case, the CFD code predicted very similar flow field structures as the test data. In this study, a CFD simulation under single-phase flow condition was conducted for one specific condition in a thermal mixing flow test of 5x5 rod bundle with some mixing vane spacer grids. In this study, a CFD simulation under a single-phase flow condition was conducted for one specific condition in a thermal mixing flow test of 5x5 rod bundle with the mixing vane spacer grids to verify the applicability of the CFD model for predicting the outlet temperature distribution. FLUENT 14.5 Version was used in this CFD analysis. For the successful prediction of the wall bounded turbulent flows, the y+ with 3 prism layers was determined within 5. At this time, k-epsilon standard turbulence model was used. The temperature distribution of CFD for each sub-channel at the outlet region of test bundle showed the difference approximately within 1.1% and 0.2% while comparing to that of test and sub-channel analysis code, respectively

  6. CFD Verification of 5x5 Rod Bundle with Mixing Vane Spacer Grids

    Park, Sungkew; Jang, Hyungwook; Lim, Jongseon; Park, Eungjun; Nahm, Keeyil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Results of the CHF test are used for determining the CHF correlation, which is used to evaluate the thermal margin in the reactor core. Computational fluid dynamics (CFD) has been used to save the time and cost for experimental tests, components design and complicated phenomena in all industries including the reactor coolant system. L. D. Smith et al. applied the CFD methodology in a 5x5 rod bundle with the mixing vane spacer grid using the renormalization group (RNG) k-epsilon model. This CFD model agreed reasonably well with the test data. M. E. Conner et al. conducted experiments to validate the CFD methodology for the single-phase flow conditions in PWR fuel assemblies. In this validation case, the CFD code predicted very similar flow field structures as the test data. In this study, a CFD simulation under single-phase flow condition was conducted for one specific condition in a thermal mixing flow test of 5x5 rod bundle with some mixing vane spacer grids. In this study, a CFD simulation under a single-phase flow condition was conducted for one specific condition in a thermal mixing flow test of 5x5 rod bundle with the mixing vane spacer grids to verify the applicability of the CFD model for predicting the outlet temperature distribution. FLUENT 14.5 Version was used in this CFD analysis. For the successful prediction of the wall bounded turbulent flows, the y+ with 3 prism layers was determined within 5. At this time, k-epsilon standard turbulence model was used. The temperature distribution of CFD for each sub-channel at the outlet region of test bundle showed the difference approximately within 1.1% and 0.2% while comparing to that of test and sub-channel analysis code, respectively.

  7. Numerical Analysis for IFM Grid Effect on 5x5 Rods Bundle

    Kim, Seong Jin; Cha, Jeong Hun; Seo, Kyong Won; Kim, Tae Woo; Kwon, Hyuk; Hwang, Dae Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    Generally, the fuel assembly consists of fuel rods, bottom and top grids, spacer grids, mixing vane, etc. The mixing vane with spacer grid is used to increase the thermal mixing between subchannels and to increase CHF(Critical Heat Flux). IFM(Intermediate Flow Mixer) grids are used to induce lateral flow between adjacent channels and are well-known as improving CHF, also. A numerical analysis using CFD code(ANSYS CFX, version 12.1) and subchannel code(MATRA-S) was conducted to investigate the influence of IFM grid on the subchannel temperature in 5x5 rods bundle with and without the IFM grid, thermohydraulically. In this study, the quantitative improvement of the mixing effect of the IFM grid is presented from the results of CFX and MATRA-S code. Moreover, capacity of predicting subchannel temperature of MATRA-S code is compared with CFX result

  8. Numerical Analysis for IFM Grid Effect on 5x5 Rods Bundle

    Generally, the fuel assembly consists of fuel rods, bottom and top grids, spacer grids, mixing vane, etc. The mixing vane with spacer grid is used to increase the thermal mixing between subchannels and to increase CHF(Critical Heat Flux). IFM(Intermediate Flow Mixer) grids are used to induce lateral flow between adjacent channels and are well-known as improving CHF, also. A numerical analysis using CFD code(ANSYS CFX, version 12.1) and subchannel code(MATRA-S) was conducted to investigate the influence of IFM grid on the subchannel temperature in 5x5 rods bundle with and without the IFM grid, thermohydraulically. In this study, the quantitative improvement of the mixing effect of the IFM grid is presented from the results of CFX and MATRA-S code. Moreover, capacity of predicting subchannel temperature of MATRA-S code is compared with CFX result

  9. Minimization of PWR reactor control rods wear

    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

  10. Minor actinide transmutation on PWR burnable poison rods

    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 keff 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 keff markedly. The PWR keff 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

  11. Tests of countercurrent flow in a 5x5 unheated road bundle

    With the objective of examining the countercurrent flow phenomenon, it was built at CDTN an Air-Water Test Facility with a 5x5 unheated rod bundle, one perfurated plate ( 'tie plate') and an plenum region, representing, in scale, the upper part of a PWR (KWU). A series of 179 tests was performed with downward injection through the upper plenum in 4 different elevations in order to investigate the effect of the injection position on the ccountercurrent flow. The phenomenon has shown a different behavior when one compares test results of the injection in the tie plate's immediate neighborhood with the remaning ones. (author)

  12. Calculation of drop course of control rod assembly in PWR

    The validation of control rod drop performance is an important part of safety analysis of nuclear power plant. Development of computer code for calculating control rod drop course will be useful for validating and improving the design of control rod drive line. Based on structural features of the drive line, the driving force on moving assembly was analyzed and decomposed, the transient value of each component of the driving force was calculated by choosing either theoretical method or numerical method, and the simulation code for calculating rod cluster control assembly (RCCA) drop course by time step increase was achieved. The analysis results of control rod assembly drop course calculated by theoretical model and numerical method were validated by comparing with RCCA drop test data of Qinshan Phase Ⅱ 600 MW PWR. It is shown that the developed RCCA drop course calculation code is suitable for RCCA in PWR and can correctly simulate the drop course and the stress of RCCA. (authors)

  13. Fuel rod behavior of a PWR during load following

    The behavior of a PWR fuel rod when operating in normal power cycles, excluding in case of accidents, is analysed. A computer code, that makes the mechanical analysis of the cladding using the finite element method was developed. The ramps and power cycles were simulated suposing the existence of cracks in pellets when the cladding-pellet interaction are done. As a result, an operation procedure of the fuel rod in power cycle is recommended. (E.G.)

  14. Control rod for PWR type reactor

    Since a silver-indium-cadmium alloy has been used as the absorber for control rods, swelling due to neutron absorption has been caused. On the other hand, a stainless steel cladding tube for the absorber gradually reduces its outer diameter by the pressure of reactor coolants and neutron irradiation and causes contact during working life to often bring about cracking in the cladding tube. Then, the control rod is divided into two independent portions and joined by an intermediate end plug into a single rod, in which the upper portion is made free from pressure and the lower portion is pressurized. Further, a large gap is formed between the lower absorber and the lower cladding tube. Further, chromium or chromium carbide is coated to the outer surface of the upper cladding tube for improving the abrasion resistance. Thus, the cladding tube is made abrasion resistant and it is possible to prevent cracking in the cladding tube due to interaction between the tube and the absorber, inner presurization at the lower portion, reduced diameter for the absorber and the gap of the tube. (N.H.)

  15. Estimating PWR fuel rod failures throughout a cycle

    A fuel performance engineer requires good prediction models for fuel conditions to help assure that any fuel repair operation he may recommend for the next refueling outage will have a minimal impact on nuclear plant operation. For nearly two decades, simple equilibrium equations have been used to provide estimates of the number of failed fuel rods in a pressurized water reactor (PWR) core. The unknown parameter is the isotopic escape rate (upsilon), which is often assumed to be --1 X 10/sup -8//s for the release of /sup 131/I from a 3- to 4-m-long PWR rod. The use of this escape rate value will generally produce end-of-cycle (EOC) predictions that are accurate within a factor of --3. When applied at the time when fuel rods initially fail, such as early in a reactor cycle, however, the prediction obtained may overestimate the number of failed rods present by a factor of 10 or more. While a goal of Combustion Engineering's (C-E's) efforts on failed fuel prediction (FFP) models over the past decade has been to increase the accuracy of the EOC estimate, recent efforts have emphasized improving prediction capability for failed rods present early in a reactor cycle. The C-E approach to modeling iodine release from failed fuel rods is based on dynamic escape rate theory that is incorporated in the C-E IODYNE (for iodine dynamic evaluation) code. This theory has been empirically modified to account for specific observed time dependencies of the release rates for /sup 131/I and /sup 133/I from a failed rod. In a current version of IODYNE, four such factors have been included in the FFP model, as described in this paper

  16. Numerical investigation on flow and heat transfer of peripheral rods in PWR rod bundle assembly

    A numerical investigation of peripheral rods in PWR rod bundle assemblies was performed to better understand the impact on flow patterns and rod heat transfer downstream of spacer grids in the peripheral rod region. Computational Fluid Dynamic (CFD) models were prepared for the peripheral rod region of typical spacer grids with and without mixing vanes to understand how grid design may have an impact on peripheral rod heat transfer. The results of the CFD analysis indicate a large degradation in heat transfer on the outer face of the peripheral rods due to increased hydraulic resistance of the grid perimeter strip and tabs for the grid without mixing vanes. With mixing vanes on the interior region of the spacer grid more flow is redistributed to the outer region of the peripheral rods so there is improved peripheral rod heat transfer. The degradation in heat transfer on the outer face of the peripheral rods for grids without mixing vanes may be a contributing factor in increasing local hotspots and boiling to increase corrosion thickness and crud deposition. To verify the improvement in peripheral rod heat transfer due to the addition of mixing vanes, corrosion data is presented from in-reactor oxide measurements for fuel with and without mixing vanes in symmetric core positions. (author)

  17. Experiments on the load following behaviour of PWR fuel rods

    KWU had studied the effects of load following operation on fuel performance from the beginning of commercial operation of nuclear power plants: The first power cycling experiments were started in 1970 in the nuclear power plant Obrigheim (KWO) and in the High Flux Reactor (HFR) Petten. These power cycling tests performed at various power levels and burnups of up to 25 GWd/t(U) showed that the fuel rod cycling performance compares well with the performance of fuel rods operated under essentially constant load at comparable power levels. Two additional cycling tests as described in this paper were performed on the HFR Petten with preirradiated PWR fuel rods having burnups of up to 40 GWd/t(U). These experiments comprised up to 60 cycles between 250/360 W/cm and 215/320 W/cm with 10% power overshoot (400, 370 W/cm) after each cycle. Also, these experiments ended up with sound fuel rods. Moreover, detailed investigations before and between power cycles and after experiment termination showed clearly that the fuel performance corresponds to a single ramp to peak power and that the cycling effects are indeed very small. This confirmed earlier findings that due to crack reversal in the UO2 the cyclic dimensional changes mainly occur in the UO2 itself. Altogether the experiments show that power cycling does not lead to fuel rod failures, which is also confirmed by successful load follow operation in commercial power plants. (orig.)

  18. Toward an early detection of PWR control rod anomalous dropping

    Some anomalous PWR control rods dropping occurred in the past. It is assumed to be caused by a geometrical deformation of its guide tube, which might be related with neutron fluence and its sharp changes. Now at days, this problem is an open field of research, oriented to the understanding and prevention of the event. Work here is focused toward early detection. A differential equation modelling control rod free fall movement is found. There result three acceleration terms: gravity; friction with fluid; and friction with its guide tube. From recorded Plant measurements, both friction coefficients are estimated. The one from guide tube experiences a large variation in case of anomalous dropping; so relationship with neutron fluence is proposed for the prevention purpose. (Author)

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

    Hursin, M.; Kochunas, B.; Downar, T. J. [Univ. of California at Berkeley, Berkeley (Canada)

    2008-10-15

    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

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

    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. Delineamento (1/5 (5 x 5 x 5 em blocos Designs (1/5 (5 x 5 x 5 in blocks

    Armando Conagin

    1982-01-01

    Full Text Available No presente trabalho, os tratamentos do delineamento fatorial fracionado (1/5 (5x5x5, obtido pela superposição de três quadrados latinos ortogonais, são colocados em cinco blocos, com a utilização de um quarto quadrado latino ortogonal. Um modelo quadrático em X foi usado para estudo da superfície de resposta, sendo considerados polinômios ortogonais linear e quadrático para cada um dos fatores e para blocos, uma vez que, em ensaios de campo, a maior parte do gradiente de fertilidade ou de outras causas sistemáticas pode ser eliminada com a estimação desses dois efeitos; foram ainda colocadas no modelo as interações lineares de dois fatores. Somente os efeitos lineares são estimados independentemente, e foram dadas, para cada fator e para blocos, as matrizes para cálculo dos efeitos quadráticos ajustados. Quando é eliminada do modelo uma das interações de dois fatores, o efeito quadrático do fator restante passa a ser estimado independentemente. Se o quarto índice for utilizado como outro fator, tem-se o delineamento (1/25 (5 x 5 x 5 x 5, completamente casualizado; este permite o estudo simultâneo de quatro fatores em cinco níveis, com apenas vinte e cinco pontos experimentais; o modelo contém efeitos lineares e quadráticos dos quatro fatores e as interações lineares desses fatores dois a dois. Se nos delineamentos (1/51 (5 x 5 x 5, divididos em cinco blocos, e (1/25 (5x5x5x5 completamente casualizado, todas as interações de dois fatores forem não-significativas, o modelo ficará só com os termos lineares e quadráticos puros, e estes poderão ser estimados independentemente, à semelhança do que ocorre com o (1/5 (5x5x5 completamente casualizado.Statistical solutions for quadratic and square root polynomials of second order for a group of (1/5 (5x5x5 fractional factorials when the design is completely randomized is briefly considered in this text. The extension of the fractional factorial (1/5 (5x5x5 to a type

  2. SIVAR - Computer code for simulation of fuel rod behavior in PWR during fast transients

    Fuel rod behavior during a stationary and a transitory operation, is studied. A computer code aiming at simulating PWR type rods, was developed; however, it can be adapted for simulating other type of rods. A finite difference method was used. (E.G.)

  3. Investigation of the load change behaviour of PWR- and BWR fuel rods at positive power ramps

    The following irradiation experiments have been performed to determine the operational behaviour of fuel rods in LWR during power ramps: a) power ramp experiment in the nuclear power plant of Obrigheim (KWO) with 6 PWR test fuel rods at a burnup of about 14 MWd/kgU. No fuel rod defects have been found. b) preirradiation of 45 segmented fuel rods in KWO and of 8 segmented fuel rods in the reactor of Wuergassen; the preirradiated segments will be ramped at HFR Petten. c) power ramp experiments at HBWR with 8 BWR test fuel rods at burnups of 4-14 MWd/kgU; ramping caused no defects. (orig.)

  4. Development of fission gas sampling system for PWR spent nuclear fuel rods and test evaluation using dummy nuclear fuel rods

    Fission gas sampling system for measuring the fission gas quantity and internal pressure of PWR spent nuclear fuel rods was developed in KAERI. This system has the advantages of reducing the time required in equilibrium pressure by using as positive pressure in the chamber when the fission gas is expanded from the fuel rod to the puncturing and standard chamber, also improving the accuracy in measuring the fuel rod internal pressure. As a results of performance evaluation test using several dummy fuel rods in the inactive region, the accuracy of measuring system appeared to be good agreement within ±5% error range

  5. Flow induced vibration analysis for preventing PWR fuel rods from excessive fretting wear

    In order to prevent PWR fuel rods excessive fretting wear, the author analysed flow induced vibration. The methods developed and used by FRAMATOME to analyze and to justify the fuel rod behaviour with respect to flow induced vibrations and wear at grid support locations were presented

  6. Effects of cladding and pellet variables on PWR fuel rod performance

    Two standard 15 x 15 PWR fuel assemblies containing test fuel rods were irradiated to an average burnup of 24,500 MWD/MTU through two cycles of operation. The assemblies had a total of 56 experimental fuel rods representing four different cladding types and two different fuel pellet types in rods located in peripheral positions. Sixteen of these test rods, representing all eight cladding/pellet combinations, were extracted from one of these assemblies for extensive nondestructive examination in the B and W LRC Hot Cells. The results obtained thus far indicate significant differences in cladding deformation and fuel pellet densification

  7. The effects of fission gas release on PWR fuel rod design and performance

    The purpose of this investigation was to determine the effects of fission gas release on PWR fuel rod design and performance. Empirical models were developed from fission gas release data. Fission gas release during normal operation is a function of burnup. There is little additional fission gas release during anticipated transients. The empirical models were used to evaluate Westinghouse fuel rod designs. It was determined that fission gas release is not a limiting parameter for obtaining rod average burnups in the range of 50 000 to 60 000 MWD/MTU. Fission gas release during anticipated transients has a negligible effect on the margins to rod design limits. (author)

  8. The effects of fission gas release on PWR fuel rod design and performance

    The purpose of this investigation was to determine the effects of fission gas release on PWR fuel rod design and performance. Empirical models were developed from fission gas release data. Fission gas release during normal operation is a function of burnup. There is little additional fission gas release during anticipated transients. The empirical models were used to evaluate Westinghouse fuel rod designs. It was determined that fission gas release is not a limiting parameter for obtaining rod average burnups in the range of 50,000 to 60,000 MWD/MTU. Fission gas release during anticipated transients has a negligible effect on the margins to rod design limits. (author)

  9. Study on the improved evaluation of radioactivity of activated control rods in PWR

    The evaluation method of radioactivity of activated materials has been developed as ORIGEN code. However, it is difficult to precisely evaluate the radioactivity of neutron absorption materials such as control rods. A control rod in PWR is made of Ag-In-Cd alloy that absorbs neutron greatly and the thermal neutron flux decreases rapidly in and around it. This phenomenon is called depression effect. The consideration of depression effect is necessary to evaluate radioactivity of the control rod. In this study we improved the reliability of the cross-section value of Ag-107(n,γ) Ag-108m by the irradiation examination in JRR3. In addition, we calculated (1) the neutron spectrum and neutron flux with depression effect by MCNP of Monte Carlo method and (2) the radioactivity of the activated control rod. The pieces of control rod were irradiated at JMTR of JAERI. As a result of the accuracy of the measurement data calculation results, we developed the method of evaluation for the radioactivity of activated control rod. The radioactivity of activated control rod in PWR was evaluated and compared with the measurement data, resulting in positive accuracy. Of special significance was confirmation of the value of Ag-108m, as an essential nuclide for long term dose estimation of disposal facility. The cross-section value of Ag-107(n,γ) Ag-108m was about one forty of existent library. This method was accurately confirmed and developed for evaluating activated control rods reasonably. (author)

  10. Axial gas flow in irradiated PWR fuel rods

    Transient and steady state axial gas flow experiments were performed on six irradiated, commercial pressurized water reactor fuel rods at ambient temperature and 533 K. Laminar flow equations, as used in the FRAP-T2 and SSYST fuel behavior codes, were used with the gas flow results to calculate effective fuel rod radial gaps. The results of these analyses were compared with measured gap sizes obtained from metallographic examination of one fuel rod. Using measured gap sizes as input, the SSYST code was used to calculate pressure drops and mass fluxes and the results were compared with the experimental gas flow data

  11. Determination of PWR control rods reactivity worth using the analysis of the power signal during a rod drop

    The present work is related to the establishment of a validation file concerning a method for determining control rod efficiency. This method is based on the analysis of the power signal obtained during a rod drop in pressurized water reactor (PWR). The main purpose of our investigations was, in particular, to analyse the various conditions which would permit large scale application in PWRS, of the above method. Five different series of measurements were first interpreted to elucidate a theoretical model. This model, in turn, was used to establish the criteria for precision and good representation. As a result of our analysis, it may now be concluded that the power signal method can estimate the total control rod efficiency (normalized to the mean power) within 5%, provided that value of total β is known. This is achieved by taking as a representative signal, the weighted sum of 8 excore detectors and five suitabily placed in core detectors

  12. The thermo-mechanics of the PWR fuel rod

    The fuel rod mechanics is of a great importance in the safety and performance of the reactors. In this domain a meeting has been organized by the SFEN the 18 march 1998 at Paris. With the participation of scientists from CEA, EDF and Framatome, the physics of the fuel rods was presented based on four main aspects. Two first papers dealt with the solicitations of the fuel rod in normal and accidental conditions. The physical phenomena under irradiation were then detailed in the four following talks. Three papers presented the simulation and the codes of the fuel-cladding interactions with the diabolo effect. The last paper was devoted to the experiment feedback and the research programs. (A.L.B.)

  13. Examination of Spent PWR Fuel Rods After 15 Years in Dry Storage

    Virginia Power Surry Nuclear Station Pressurized Water Reactor (PWR) fuel was stored in a dry inert atmosphere Castor V/21 cask at the Idaho National Environmental and Engineering Laboratory (INEEL) for 15 years at peak cladding temperatures decreasing from about 350 to 150 deg. C. Prior to the storage, the loaded cask was subjected to extensive thermal benchmark tests. The cask was opened to examine the fuel for degradation and to determine if it was suitable for extended storage. No rod breaches had occurred and no visible degradation or crud/oxide spallation were observed. Twelve rods were removed from the center of the T11 assembly and shipped from INEEL to the Argonne-West HFEF for profilometric scans. Four of these rods were punctured to determine the fission gas release from the fuel matrix and internal pressure in the rods. Three of the four rods were cut into five segments each, then shipped to the Argonne-East AGHCF for detailed examination. The test plan calls for metallographic examination of six samples from two of the rods, microhardness and hydrogen content measurements at or near the six metallographic sample locations, tensile testing of six samples from the two rods, and thermal creep testing of eight samples from the two rods to determine the extent of residual creep life. The results from the profilometry (12 rods), gas release measurements (4 rods), metallographic examinations (2 samples from 1 rod), and microhardness and hydrogen content characterization (2 samples from 1 rod) are reported here. The tensile and creep studies are just starting and will be reported at a later date, along with the additional characterization work to be performed. Although only limited pre-storage characterization is available, a number of preliminary conclusions can be drawn based on comparison with characterization of Florida Power Turkey Point rods of a similar vintage. Based on this comparison, it appears that little or no cladding thermal creep and fission

  14. The physical and chemical degradation of PWR fuel rods in severe accident conditions

    An experimental study of the interaction between Zircaloy-4 cladding and UO2 in PWR fuel rods heated to high temperatures with a negligible differential pressure across the cladding wall is described. The fuel rods were of dimensions appropriate to the 17x17 PWR fuel sub-assembly and were heated in a non-oxidising environment (vacuum) up to approx. 1850 deg. C either isothermally or through heating ramps. Observations were made concerning the extent and nature of the reaction zone between Zircaloy-4 and UO2 over the temperature range 1500-1850 deg. C for times ranging from 1 min to 125 min. The location, morphology and the chemical composition of the phases formed are described along with the kinetics of their formation. (author)

  15. Pressure equalization system in PWR-fuel rods

    The pressure equalization system, developed on the basis of activated charcoal, is capable of reducing the internal pressure rise in fuel rods by adsorption of the fission gases. He-prepressure does not affect the system and Helium will not be adsorbed. Irradiation does not reduce the adsorption capacity of activated charcoal down to an unacceptable limit. Shaped activated charcoal is a suitable material which can be well defined and characterized. Feasible techniques of activating and assembling methods can be proposed. (orig.)

  16. Grid to rod fretting wear in EDF PWR from operating problems to new designs qualification method

    For the last ten years, problems of leaking fuel rod occurred in PWR Power Plants throughout the world, which were due to grid to rod fretting wear. Two main reasons were identified: high flow induced vibration sensitivity or insufficient end of life rod tightening. Leaking fuel rods contaminate primary coolant. The potential consequences are: (i) anticipated plant shutdown in order to respect the mandatory fission gas threshold in the primary coolant, (ii) changes in the loading pattern, (iii) special operating personnel training and plant shutdown precautions to prevent contamination risks. For EDF, it is a safety and financial stake. EDF and its industrials partners have launched an important R and D program to provide EDF, with analytic tools and methods to prevent new design from experiencing fretting problems. This paper presents how EDF qualifies each new design proposed by the fuel vendors. EDF qualification method is based on three approaches (i) Operating feedback, (ii) Fuel assembly tests, (iii) Numeric simulation. (i) French PWR standardisation authorises to extend one plant operating feedback to the other plant with the same standard. Operating feedback gives relevant information regarding models and the qualification test methodology. (ii) Two tests are performed: vibration test to study the FA response under hydraulic loading and grid to rod fretting resistance test. (iii) Rod vibration behaviour models have been developed on the base of the EDF Finite Element code CodeAster with an equivalent linear grid to rod contact modelisation, or contact elements. For both tests and mechanic models, the boundary conditions are grid to rod contact conditions, calculated with EDF analytic model, and hydraulic loading, determined using thermo hydraulic computation and analytic models. (author)

  17. In-pile post-DNB behavior of a nine-rod PWR-type fuel bundle

    The results of an in-pile power-cooling-mismatch (PCM) test designed to investigate the behavior of a nine-rod, PWR-type fuel bundle under intermittent and sustained periods of high temperature film boiling operation are presented. Primary emphasis is placed on the DNB and post-DNB events including rod-to-rod interactions, return to nucleate boiling (RNB), and fuel rod failure. A comparison of the DNB behavior of the individual bundle rods with single-rod data obtained from previous PCM tests is also made

  18. Behaviour of a defective MOX fuel rod in a PWR

    At then end of the DAMPIERRE 1 power plant 11th cycle in 1993, routine sipping techniques identified a leak on a mixed oxide (MOX) fuel assembly. Application of failed fuel management policy led to the reloading of this assembly for the 12th cycle. It was its second irradiation campaign. This situation allowed EDF and CEA to set up a special monitoring programme during the cycle in order to obtain numerous information about the behaviour of defective MOX fuel at steady state power levels and during transients. These data have been obtained due to: a very good knowledge of its power history; the daily routine measurements of the primary circuit activity, conducted by the plant operator; a specific on-line gamma spectrometry facility installed by CEA to monitor the primary water during this cycle; various on-site post-irradiation examinations carried out at the end of the 12th cycle: visual inspection, qualitative and quantitative sipping tests, defective fuel rods detection and localization. Analyses of this behaviour were mainly in relation with two scopes: determination of the gaseous and airborne fission product release rates out of a defective MOX stack, in relation with theoretical prediction (PROFIP code) and with results obtained on experimental similar rods in a research reactor (EDITHMOX 01 experiment); application of on-line discrimination methods developed by EDF and CEA in order to know the type (uranium oxide UO2 or mixed oxide) of fission product source. Based on measurement of gaseous isotopes, these methods allow plant operator to detect very early a possible evolution of the defect and to foresee on-site examinations during shutdown period. The main results of this programme were: release rates of gaseous fission product were similar to those observed with defective UO2 fuel; no worsening of the defect size or of activity release occurred over one year of irradiation; discrimination methods have been in good agreement with the type of fuel. (author

  19. PCI performance of PWR rods with excessive oxide spalling and large hydrogen content

    It is well known that hydrides in Zircaloy materials may embrittle the material and deteriorate its mechanical performance. It has also been argued in the open literature that hydrides in Zircaloy materials may impact the stress corrosion cracking (SCC) mechanism resulting in pellet cladding interaction (PCI) failures at lower rod power levels. This paper describes a study to assess if, and in such a case to what degree, hydrides in Zircaloy may lower the PCI resistance. Five fuel rod segments were manufactured from 2 PWR Zr-4 rods irradiated to a rod average burnup of about 35 MWd/kgU. The segments showed excessive oxide spalling resulting in local concentration of the cladding hydrides. Five of these segments were ramp tested in the Studsvik R-2 research reactor to power levels ranging from 45 to 50 kW/m. The paper discusses the ramp results and more specifically the potential role of hydrides/thick oxide layers on PWR fuel cladding PCI performance. (author)

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

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

    2013-07-01

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

  1. INTERCOMPARISON OF RESULTS FOR A PWR ROD EJECTION ACCIDENT

    DIAMOND,D.J.; ARONSON,A.; JO,J.; AVVAKUMOV,A.; MALOFEEV,V.; SIDOROV,V.; FERRARESI,P.; GOUIN,C.; ANIEL,S.; ROYER,M.E.

    1999-10-01

    This study is part of an overall program to understand the uncertainty in best-estimate calculations of the local fuel enthalpy during the rod ejection accident. Local fuel enthalpy is used as the acceptance criterion for this design-basis event and can also be used to estimate fuel damage for the purpose of determining radiological consequences. The study used results from neutron kinetics models in PARCS, BARS, and CRONOS2, codes developed in the US, the Russian Federation, and France, respectively. Since BARS uses a heterogeneous representation of the fuel assembly as opposed to the homogeneous representations in PARCS and CRONOS, the effect of the intercomparison was primarily to compare different intra-assembly models. Quantitative comparisons for core power, reactivity, assembly fuel enthalpy and pin power were carried out. In general the agreement between methods was very good providing additional confidence in the codes and providing a starting point for a quantitative assessment of the uncertainty in calculated fuel enthalpy using best-estimate methods.

  2. Water-side oxide layer thickness measurement of the irradiated PWR fuel rod by NDT method

    It has been known that water-side corrosion of fuel rods in nuclear reactor is accompanied with the loss of metallic wall thickness and pickup of hydrogen. This corrosion is one of the important limiting factors in the operating life of fuel rods. In connection with the fuel cladding corrosion, a device to measure the water-side oxide layer thickness by means of the eddy-current method without destructing the fuel rod was developed by KAERI. The device was installed on the multi-function testing bench in the nondestructive test hot-cell and its calibration was carried out successfully for the standard rod attached with plastic thin films whose thicknesses are predetermined. It shows good precision within about 10% error. And a PWR fuel rod, one of the J-44 assembly discharged from Kori nuclear power plant Unit-2, has been selected for oxide layer thickness measurements. With the result of data analysis, it appeared that the oxide layer thicknesses of Zircaloy cladding vary with the length of the fuel rod, and their thicknesses were compared with those of the destructive test results to confirm the real thicknesses

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

    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.)

  4. Post-irradiation examination of a bowed PWR fuel rod with contact

    During reactor operation in the Ringhals 2 PWR a fuel rod bowed and as a result came in contact with an adjacent rod. The rod contact in one of the lower grid spans was observed during visual inspection at the end of life. The visual appearance suggested that there was possibly increased cladding corrosion on both the contacting rods at and close to the position of contact. One of the contacting rods was sent to Studsvik’s Hot Cell Laboratory for investigation where fission gas analysis, gamma scanning, EC oxide thickness, metallography (optical microscopy) and cladding microhardness measurements were performed in order to verify the impact of the bow and the contact on the fuel rod performance, with particular focus on the local cladding corrosion. The influence of the reduction of moderator in the region of the contact point was seen in the Cs-137 axial gamma scanning and in the Ce-144 rotational gamma scanning, which show a local reduction of both the pellet-average power, in the contact region, and specifically on the side with the contact. Visual inspection revealed increased corrosion in the rod-rod contact position. Metallographic examination of a cross-section at the elevation with the contact showed that increased corrosion and loss of material had occurred at the contact position. Outside of the immediate vicinity of the contact region the corrosion was not affected. The cladding microhardness was measured at different radial positions both at the contact position and at other positions around the cladding circumference. Based on the relationship between the microhardness and local temperature during operation on fully wetted cladding, it was possible to estimate the cladding surface temperature at the contact point to approximately 360°C. This local overheating and conditions arising from the local overheating can explain the higher local oxidation of the cladding observed in the visual inspection and metallography. (author)

  5. A model finite-element to analyse the mechanical behavior of a PWR fuel rod

    A model to analyse the mechanical behavior of a PWR fuel rod is presented. We drew our attention to the phenomenon of pellet-pellet and pellet-cladding contact by taking advantage of an elastic model which include the effects of thermal gradients, cladding internal and external pressures, swelling and initial relocation. The problem of contact gives rise ro a variational formulation which employs Lagrangian multipliers. An iterative scheme is constructed and the finite element method is applied to obtain the numerical solution. Some results and comments are presented to examine the performance of the model. (author)

  6. The three-dimensional PWR transient code ANTI; rod ejection test calculation

    ANTI is a computer program being developed for three-dimensional coupled neutronics and thermal-hydraulics description of a PWR core under transient conditions. In this report a test example calculated by the program is described. The test example is a simulation of a control rod ejection from a very small reactor core (to save somputing time). In order to show the influence of cross flow between adjacent fuel elements the same calculation was performed both with the cross flow option and with closed hydraulic channels. (author)

  7. The Verification of Coupled Neutronics Thermal-Hydraulics Code NODAL3 in the PWR Rod Ejection Benchmark

    2014-01-01

    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 t...

  8. Composition and Distribution of Tramp Uranium Contamination on BWR and PWR Fuel Rods

    In a joint research project of VGB and AREVA NP GmbH the behaviour of alpha nuclides in nuclear power plants with light water reactors has been investigated. Understanding the source and the behaviour of alpha nuclides is of big importance for planning radiation protection measures for outages and upcoming dismantling projects. Previous publications have shown the correlation between plant specific alpha contamination of the core and the so called 'tramp fuel' or 'tramp uranium' level which is linked to the defect history of fuel assemblies and accordingly the amount of previously washed out fuel from defective fuel rods. The methodology of tramp fuel estimation is based on fission product concentrations in reactor coolant but also needs a good knowledge of tramp fuel composition and in-core distribution on the outer surface of fuel rods itself. Sampling campaigns of CRUD deposits of irradiated fuel assemblies in different NPPs were performed. CRUD analyses including nuclide specific alpha analysis have shown systematic differences between BWR and PWR plants. Those data combined with literature results of fuel pellet investigations led to model improvements showing that a main part of fission products is caused by fission of Pu-239 an activation product of U-238. CRUD investigations also gave a better picture of the in-core composition and distribution of the tramp uranium contamination. It was shown that the tramp uranium distribution in PWR plants is time dependent. Even new fuel assemblies will be notably contaminated after only one cycle of operation. For PWR applies the following logic: the higher the local power the higher the contamination. With increasing burnup the local rod power usually decreases leading to decreasing tramp uranium contamination on the fuel rod surface. This is not applicable for tramp uranium contamination in BWR. CRUD contamination (including the tramp fuel deposits) is much more fixed and is constantly increasing

  9. Rod consolidation of RG and E's [Rochester Gas and Electric Corporation] spent PWR [pressurized water reactor] fuel

    The rod consolidation demonstration involved pulling the fuel rods from five fuel assemblies from Unit 1 of RG and E's R.E. Ginna Nuclear Power Plant. Slow and careful rod pulling efforts were used for the first and second fuel assemblies. Rod pulling then proceeded smoothly and rapidly after some minor modifications were made to the UST and D consolidation equipment. The compaction ratios attained ranged from 1.85 to 2.00 (rods with collapsed cladding were replaced by dummy rods in one fuel assembly to demonstrate the 2:1 compaction ratio capability). This demonstration involved 895 PWR fuel rods, among which there were some known defective rods (over 50 had collapsed cladding); no rods were broken or dropped during the demonstration. However, one of the rods with collapsed cladding unexplainably broke during handling operations (i.e., reconfiguration in the failed fuel canister), subsequent to the rod consolidation demonstration. The broken rod created no facility problems; the pieces were encapsulated for subsequent storage. Another broken rod was found during postdemonstration cutting operations on the nonfuel-bearing structural components from the five assemblies; evidence indicates it was broken prior to any rod consolidation operations. During the demonstration, burnish-type lines or scratches were visible on the rods that were pulled; however, experience indicates that such lines are generally produced when rods are pulled (or pushed) through the spacer grids. Rods with collapsed cladding would not enter the funnel (the transition device between the fuel assembly and the canister that aids in obtaining high compaction ratios). Reforming of the flattened areas of the cladding on those rods was attempted to make the rod cross sections more nearly circular; some of the reformed rods passed through the funnel and into the canister

  10. Experimental benchmark data for PWR rod bundle with spacer-grids

    In numerical simulations of fuel rod bundle flow fields, the unsteady Navier–Stokes equations have to be solved in order to determine the time (phase) dependent characteristics of the flow. In order to validate the simulations results, detailed comparison with experimental data must be done. Experiments investigating complex flows in rod bundles with spacer grids that have mixing devices (such as flow mixing vanes) have mostly been performed using single-point measurements. In order to obtain more details and insight on the discrepancies between experimental and numerical data as well as to obtain a global understanding of the causes of these discrepancies, comparisons of the distributions of complete phase-averaged velocity and turbulence fields for various locations near spacer-grids should be performed. The experimental technique Particle Image Velocimetry (PIV) is capable of providing such benchmark data. This paper describes an experimental database obtained using two-dimensional Time Resolved Particle Image Velocimetry (TR-PIV) measurements within a 5 × 5 PWR rod bundle with spacer-grids that have flow mixing vanes. One of the unique characteristic of this set-up is the use of the Matched Index of Refraction technique employed in this investigation to allow complete optical access to the rod bundle. This unique feature allows flow visualization and measurement within the bundle without rod obstruction. This approach also allows the use of high temporal and spatial non-intrusive dynamic measurement techniques namely TR-PIV to investigate the flow evolution below and immediately above the spacer. The experimental data presented in this paper includes explanation of the various cases tested such as test rig dimensions, measurement zones, the test equipment and the boundary conditions in order to provide appropriate data for comparison with Computational Fluid Dynamics (CFD) simulations. Turbulence parameters of the obtained data are presented in order to gain

  11. Methodology of PWR fuel rod vibration and fretting evaluation in HERMES facilities

    In recent years, more demanding PWR core designs and operating strategies (longer cycle length, higher burn up, low leakage core...) have been chosen to improve capacity factors and fuel cycle economics. Moreover, in the current environment, fuel reliability remains an important concern. Within this context, fuel failures caused by grid-to-rod fretting now appear as a main issue due to new fuel assembly designs and their interaction with operating plants. In order to evaluate fuel assembly (FA) performances in terms of grid-to-rod fretting prior to its in-core operating, the CEA has developed the capacity to test fuel assembly prototypes in full-scale facilities so that wear and fuel vibrations can be evaluated experimentally. This paper will present these two steps: - The first step will consist of evaluating the fuel vibrations and fluid velocity fields in the HERMES T facility under variable flow conditions, - The second step will involve wear evaluation at nominal PWR conditions in the HERMES P facility using a 3-dimensional wear bench especially adapted to characterize fuel rod wear extensively. In order to improve FA qualification, the fuel assembly boundary conditions in operating plants have been studied and led to specific modifications in the HERMES facilities. The PWR scale 1 test facility HERMES P operating at nominal conditions has been modified to take these evolutions into account owing to our improved knowledge of both the mechanical and hydraulic in-reactor FA behavior. One of the major improvements, that is to be presented in the paper, concerns the hydraulic loop which has been modified. Its purpose is to simulate various in-core redistributions. Along with scale 1 test facilities, some 'reduced scale' test facilities (GRILLON) are operated to test determining parameters (temperature, test duration, etc...). Analytical results have been used as input data to define the methodology for the wear experimentation in test facilities of fuel assembly

  12. Nondestructive testing of PWR type fuel rods by eddy currents and metrology in the OSIRIS reactor pool

    The Saclay Reactor Department has developed a nondestructive test bench, now installed above channel 1 of the OSIRIS reactor. As part of investigations into the dynamics of PWR fuel degradation, a number of fuel rods underwent metrological and eddy current inspection, after irradiation

  13. Reproducibility of heat transfer tests in a 5X5 bundle geometry

    This paper describes the repeatability and reliability of bundle heat transfer data obtained in a 5X5 PWR-type bundle subassembly operating at PWR conditions of interest. The 5X5 fuel bundle simulator, installed in the OMEGA-2 loop, is equipped with simple support grids, designed to have a low impact on the flow and heat transfer. The nine central heaters were equipped with the novel sliding thermocouple technique, capable of measuring the detailed axial and circumferential temperature distributions during single-phase and boiling heat transfer tests. In order to obtain highly accurate bundle heat transfer measurements, appropriate experimental procedures and in-situ calibrations of all essential instrumentation were employed. This includes (i) the employment of calibrated reference fluid temperature measurement devices, (ii) in-situ calibrations of fluid and heater-sheath thermocouples, (iii) calibration of heater wall thickness based on in-situ measurements, and (iv) selection of data that satisfy strict acceptance criteria. After applying these corrections and data screening criteria, the measurement accuracy and repeatability was assessed. This was done by means of three different tests: Single Phase Heat Transfer: The repeatability of heat transfer were assessed by comparing the measurements of two separate 5X5 bundles against the predictions from a Dittus-Boelter-type heat transfer correlation which provided very similar results. Also the single-phase heat transfer repeatability was assessed by performing several repeat runs and comparing results obtained on heaters in symmetric locations. Excellent repeatability was noted and the results for symmetric angular locations are almost identical; Boiling Tests: During the boiling heat transfer tests excellent repeatability and symmetry was observed. The saturation temperature (corresponding to the measured outlet temperature) was found to be in very good agreement with (i) the outlet temperature measured by the

  14. Parametric Study of Control Rod Exposure for PWR Burnup Credit Criticality Safety Analyses

    The Interim Staff Guidance on burnup credit (ISG-8) for pressurized water reactor (PWR) spent nuclear fuel (SNF), issued by the Nuclear Regulatory Commission's (NRC) Spent Fuel Project Office, recommends the use of analyses that provide an ''adequate representation of the physics'' and notes particular concern with the ''need to consider the more reactive actinide compositions of fuels burned with fixed absorbers or with control rods fully or partly inserted.'' In the absence of readily available information on the extent of control rod (CR) usage in U.S. PWRs and the subsequent reactivity effect of CR exposure on discharged SNF, NRC staff have indicated a need for greater understanding in these areas. In response, this paper presents results of a parametric study of the effect of CR exposure on the reactivity of discharged SNF for various CR designs (including Axial Power Shaping Rods), fuel enrichments, and exposure conditions (i.e., burnup and axial insertion). The study is performed in two parts. In the first part, two-dimensional calculations are performed, effectively assuming full axial CR insertion. These calculations are intended to bound the effect of CR exposure and facilitate comparisons of the various CR designs. In the second part, three-dimensional calculations are performed to determine the effect of various axial insertion conditions and gain a better understanding of reality. The results from the study demonstrate that the reactivity effect increases with increasing CR exposure (e.g., burnup) and decreasing initial fuel enrichment (for a fixed burnup). Additionally, the results show that even for significant burnup exposures, minor axial CR insertions (e.g., eff of a spent fuel cask

  15. Temperature escalation in PWR fuel rod simulators due to the zircaloy/steam reaction: Tests ESSI-1,2,3

    This report discusses the test conduct, results, and posttest appearance of three scoping tests (ESSI-1,2,3) investigating temperature escalation in zircaloy clad fuel rods. The experiments are part of an out-of-pile program using electrically heated fuel rod simulators to investigate PWR fuel element behavior up to temperatures of 20000C. These experiments are part of the PNS Severe Fuel Damage Program. The temperature escalation is caused by the exothermal zircaloy/steam reaction, whose reaction rate increases exponentially with the temperature. The tests were performed using different initial oxide layers as a major parameter, obtained by varying the heatup rates and steam exposure times. (orig./RW)

  16. Study of development of non-destructive method for determining FGR from high burned PWR type fuel rod

    Experimental study was made to evaluate the FGR (Fission Product Gas Release) from high burned PWR type fuel rods by means of non-destructive method through measurement of the gamma activity of 85Kr isotope which was accumulated in the fuel top plenum. Experimental result shows that it is possible to know the amounts of FGR at fuel plenum by the equations given in the followings. FGR = 0.28C/Vf or FGR = 0.07C where, FGR (%) is the amounts of Xe and Kr released from UO2 fuel, C (counts/h) the radioactivity of 85Kr at plenum of the tested fuel rod and Vf (ml) the plenum volume of the tested fuel rod, respectively. The present study was made by using 14 x 14 PWR type fuel rods preirradiated up to the burn-up of 42.1 MWd/kgU, followed by the pulse irradiation at Nuclear Safety Research Reactor of Japan Atomic Energy Research Institute (JAERI). The FGR of the tested segmented fuel rods were measured by puncturing and found to range from 0.6% to 12% according to the magnitude of the deposited energy given by pulse. Estimated experimental error bands against the above equations were within plus minus 30%. (author)

  17. The Verification of Coupled Neutronics Thermal-Hydraulics Code NODAL3 in the PWR Rod Ejection Benchmark

    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.

  18. PWR Control Rod Ejection Analysis with the Method Of Characteristic Code DeCART

    Hursin, Mathieu; Downar, Thomas J. [University of California at Berkeley, Berkeley (United States); Thomas, Justin [Argonne National Laboratory, Argonne (United States)

    2008-07-01

    During the past several years, a comprehensive high fidelity reactor core modeling capability has been developed called the Numerical Nuclear Reactor (NNR) (Weber,2003) for detailed analysis of Light Water Reactors. The NNR achieves high fidelity with a whole-core neutron transport solution and ultra-fine-mesh computational fluid dynamics/heat transfer solution. Previous applications of the NNR have been to the steady-state analysis of both pressurized and boiling water reactors. Recently there has been interest in taking advantage of the NNR to improve the fidelity for PWR transient analysis. The work described in this paper is a preliminary demonstration of the ability of the whole core neutron transport code, DeCART, to provide a detailed intra-pin-power distribution during a control rod ejection accident. The current state of the art in analysis of this event relies upon the assembly averaged power from a whole core nodal neutronics simulator and some type of pin power reconstruction within the fuel assembly. Both methodologies are briefly presented and applied to model a super-prompt reactivity insertion accident. The difference in the results of both approaches are discussed and the benefit of the DeCART methodology is described. (authors)

  19. PWR Control Rod Ejection Analysis with the Method Of Characteristic Code DeCART

    During the past several years, a comprehensive high fidelity reactor core modeling capability has been developed called the Numerical Nuclear Reactor (NNR) (Weber,2003) for detailed analysis of Light Water Reactors. The NNR achieves high fidelity with a whole-core neutron transport solution and ultra-fine-mesh computational fluid dynamics/heat transfer solution. Previous applications of the NNR have been to the steady-state analysis of both pressurized and boiling water reactors. Recently there has been interest in taking advantage of the NNR to improve the fidelity for PWR transient analysis. The work described in this paper is a preliminary demonstration of the ability of the whole core neutron transport code, DeCART, to provide a detailed intra-pin-power distribution during a control rod ejection accident. The current state of the art in analysis of this event relies upon the assembly averaged power from a whole core nodal neutronics simulator and some type of pin power reconstruction within the fuel assembly. Both methodologies are briefly presented and applied to model a super-prompt reactivity insertion accident. The difference in the results of both approaches are discussed and the benefit of the DeCART methodology is described. (authors)

  20. Towards a reference numerical scheme using MCNPX for PWR control rod tip fluence estimations

    Recent occurrences of cracks and fissures on the cladding tubes of PWR control rod (CR) fingers employed in the Swiss reactors prompted the need to develop more reliable analytical methods for CR tip fluence estimations. To partly address this need, a deterministic methodology based on SIMULATE-3/CASMO-4 was in recent years developed at PSI. Although this methodology has already been applied for independent support to licensing issues related to CR lifetime, two main questions are currently being the center of attention for further enhancements. First, the methodology relies on several assumptions that have so far not been verified. Secondly, an assessment of the achieved accuracy has not been addressed. In an attempt to answer both these open questions, it was considered appropriate to develop an alternative computational scheme based on the stochastic MCNPX code with the objective to provide reference numerical solutions. This paper presents the first steps undertaken in that direction. To start, a methodology for a volumetric neutron source transfer to full core MCNPX models with detailed CR as well as axial reflector representations is established. On this basis, the assumptions of the deterministic methodology are studied for selected CR configurations for two Beginning-of-Life cores by comparing the spatial neutron flux distributions obtained with the two approaches for the entire spectrum. Finally, for the high-energy range (E> 1 MeV) and for a few CRs, the new MCNPX scheme is applied to estimate the accumulated fluence over one real operated cycle and the results are compared with the deterministic approach. (authors)

  1. Modeling of the PWR fuel mechanical behaviour and particularly study of the pellet-cladding interaction in a fuel rod

    In Pressurized Water Reactor (PWR) power plants, fuel cladding constitutes the first containment barrier against radioactive contamination. Computer codes, developed with the help of a large experimental knowledge, try to predict cladding failures which must be limited in order to maintain a maximal safety level. Until now, fuel rod design calculus with unidimensional codes were adequate to prevent cladding failures in standard PWR's operating conditions. But now, the need of nuclear power plant availability increases. That leads to more constraining operating condition in which cladding failures are strongly influenced by the fuel rod mechanical behaviour, mainly at high power level. Then, the pellet-cladding interaction (PCI) becomes important, and is characterized by local effects which description expects a multidimensional modelization. This is the aim of the TOUTATIS 2D-3D code, that this thesis contributes to develop. This code allows to predict non-axisymmetric behaviour too, as rod buckling which has been observed in some irradiation experiments and identified with the help of TOUTATIS. By another way, PCI is influenced by under irradiation experiments and identified with the help of TOUTATIS which includes a densification model and a swelling model. The latter can only be used in standard operating conditions. However, the processing structure of this modulus provides the possibility to include any type of model corresponding with other operating conditions. In last, we show the result of these fuel volume variations on the cladding mechanical conditions. (author). 25 refs., 89 figs., 2 tabs., 12 photos., 5 appends

  2. Post test investigation of the single rod tests ESSI 1-11 on temperature escalation in PWR fuel rod simulators due to the Zircaloy/steam reaction

    This KfK-report describes the posttest investigation of the single rod tests ESSI-1 to ESSI-11. The objective of these tests was to investigate the temperature escalation behaviour of Zircaloy clad PWR-fuel rods in steam. The investigation of the temperature escalation is part of the program of out-of-pile experiments (CORA) performed within the frame work of the PNS Severe Fuel Damage Program. The experimental arrangement consisted of fuel rod simulator (central tungsten heater, UO2 ring pellets and Zircaloy cladding), Zircaloy shroud and fiber ceramic insulation. The introductory test ESSI-1 to ESSI-3 were scoping tests designed to obtain information on the temperature escalation of zircaloy in steam. ESSI-4 to ESSI-8 were run with increasing heating rates to investigate the influence of the oxide layer thickness at the start of the escalation. ESSI-9 to ESSI-11 were performed to investigate the influence of the insulation thickness on the escalation behaviour. In these tests we also learned that the gap between removed shroud and insulation has a remarkable influence due to heat removal by convection in the gap. After the test the fuel rod simulator was embedded into epoxy and cut by a diamond saw. The cross sections were photographed and investigated by metalograph microscope, SEM and EMP examinations. (orig./GL)

  3. Analysis of high burnup fuel behavior under rod ejection accident in the Westinghouse-designed 950 MWe PWR

    As there has arisen a concern that failure of the high burnup fuel under the reactivity-insertion accident (RIA) may occur at the energy lower than the expected, duel behavior under the rod ejection accident in a typical Westinghouse-designed 950 MWe PWR was analyzed by using the three dimensional nodal transient neutronics code, PANBOX2 and the transient fuel rod performance analysis code, FRAP-T6. Fuel failure criteria versus the burnup was conservatively derived taking into account available test data and the possible fuel failure mechanisms. The high burnup and longer cycle length fuel loading scheme of a peak rod burnup of 68 MWD/kgU was selected for the analysis. Except three dimensional core neutronics calculation, the analysis used the same core conditions and assumptions as the conventional zero dimensional analysis. Results of three dimensional analysis showed that the peak fuel enthalpy during the rod ejection accident is less than one third of that calculated by the core is less than 4 percent. Therefore, it can be said that the current design limit of less than 10 percent fuel failure and maintaining the core coolable geometry would be adequately satisfied under the rod ejection accident, even though the conservative fuel failure criteria derived from the test data are applied. (author)

  4. A Calculation of the radioactivity induced in PWR cluster control rods with the origin and casmo codes

    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)

  5. Modelling of CRUD growth phenomena on PWR fuel rods under nucleate boiling conditions

    PWR primary circuit materials undergo general corrosion leading to a release of metallic element release and subsequent process of particle deposition and ion precipitation on the primary circuit surfaces. The species accumulated on fuel rods are activated by neutron flux. Consequently, crud erosion and dissolution induce primary coolant contamination. In French PWRs, 58Co volume activity is generally low and almost constant (< 30 MBq.m-3) throughout an ordinary operating cycle. In some specific cases, a significant increase in volume activity is observed after the middle of a cycle (100-1000 MBq.m-3 for 58Co) when conditions for nucleate boiling are locally reached in certain fuel assemblies. Indeed, it is well known that nucleate boiling intensifies the deposition process. The thickness of the crud layer can reach some micrometers in non-boiling areas, whereas it can reach 100 micrometers in boiling areas. Crud growth in boiling conditions can be related to three phenomena: bubble growth induces deposition process (called boiling deposition), bubbles induce concentration increase at crud-coolant interface (called enrichment and modelled by the enrichment factor, the ratio between the wall concentration and the bulk concentration) and vaporisation induces concentration increase inside the crud. A literature review on the modelling of these phenomena and on the crud structure in nucleate boiling conditions has been carried out. The OSCAR [1] calculation code developed by the CEA to predict surface and volume activities in a single phase PWR primary circuit was chosen as a basis for present study. Ability to describe local nucleate boiling conditions was added to this code leading to realistic modelling of subsequent volume activity increase. In this article, we present the results obtained using a modified version of the OSCAR PC V1.2 calculation code including: - A double phase thermal-hydraulic module, - A model of boiling crud growth, able to calculate inner

  6. Study of transient heat transfer in a fuel rod 3D, in a situation of unplanned shutdown of a PWR

    Affonso, Renato Raoni Werneck; Martins, Rodolfo Ienny; Sampaio, Paulo Augusto Berquo de; Moreira, Maria de Lourdes, E-mail: raoniwa@yahoo.com.br, E-mail: rodolfoienny@gmail.com, E-mail: sampaio@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The study, in situations involving accidents, of heat transfer in fuel rods is of known importance, since it can be used to predict the temperature limits in designing a nuclear reactor, to assist in making more efficient fuel rods, and to increase the knowledge about the behavior of the reactor's components, a crucial aspect for safety analysis. This study was conducted using as parameter the fuel rod that has the highest average power in a typical PWR reactor. For this, we developed a program (Fuel{sub R}od{sub 3}D) in Fortran language using the Finite Elements Method (FEM) for the discretization of a fuel rod and coolant channel, in order to study the temperature distribution in both the fuel rod and the coolant channel. Transient parameters were coupled to the heat transfer equations in order to obtain details of the behavior of the rod and the channel, which allows the analysis of the temperature distribution and its change over time. This work aims to present a study case of an accident where there is a lack of energy in the reactor's coolant pumps and in the diesel engines, resulting in an unplanned shutdown of the reactor. In order to achieve the intended goal, the present work was divided as follows: a short introduction about heat transfer, including the equations concerning the fuel rod and the energy equation in the channel, an explanation about how the verification of the Fuel{sub R}od{sub 3}D program was made, and the analysis of the results. (author)

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

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

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

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

    1983-10-01

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

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

    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)

  10. Process development and fabrication for sphere-pac fuel rods. [PWR; BWR

    Welty, R.K.; Campbell, M.H.

    1981-06-01

    Uranium fuel rods containing sphere-pac fuel have been fabricated for in-reactor tests and demonstrations. A process for the development, qualification, and fabrication of acceptable sphere-pac fuel rods is described. Special equipment to control fuel contamination with moisture or air and the equipment layout needed for rod fabrication is described and tests for assuring the uniformity of the fuel column are discussed. Fuel retainers required for sphere-pac fuel column stability and instrumentation to measure fuel column smear density are described. Results of sphere-pac fuel rod fabrication campaigns are reviewed and recommended improvements for high throughput production are noted.

  11. SEFLEX fuel rod simulator effects in flooding experiments. Pt. 3

    This report presents typical data and a limited heat transfer analysis from blocked bundle reflood tests of an experimental thermal-hydraulic program. Full-length bundles of 5x5 fuel rod simulators having a gas-filled gap between the Zy cladding and the alumina pellets were tested in the test rig designed for the earlier Flooding Experiments with Blocked Arrays (FEBA-program). The 5x5 FEBA rod bundle tests were performed with gapless heater rods. These rods have a close thermal contact between the stainless steel cladding and the electric insulation material. A comparison of the SEFLEX data with the reference data of FEBA obtained under identical initial and reflood conditions shows the influence of different fuel rod simulators on the thermal-hydraulic behavior during forced feed bottom reflooding of unblocked and blocked arrays. Compared to bundles of gapless rods, bundles of rods with Zy claddings and a gas filled gap between claddings and pellets, which more closely represent the features that exist in an actual fuel rod geometry, produced higher quench front velocities, enhanced removal of stored heat in the rods, reduced peak cladding temperatures, increased grid spacer effects and absolutely unproblematic coolability of 90 percent blockages with bypass. The data offer the opportunity for further validation of computer codes to make realistic predictions of safety margins during a LOCA in a PWR. (orig./HP)

  12. The feasibility of using neural networks for determination of control rod elevation in a PWR

    This paper presents the results of a preliminary study on using neural networks for determination of the axial position of control rods in PWRs. The method is based on the dependence of the axial flux profile on control rod elevation in a reactor. This flux profile can be measured by e.g. a moveable detector in an operating plant. However, in this preliminary study the flux profile is only calculated using an advanced core code for several axial positions of a partially inserted control rod. The calculated fluxes with corresponding positions of the control rod are used for training a neural network. Using the trained network it is then possible to determine the unknown axial position of a control rod elevation from the corresponding axial flux profile. 10 refs

  13. Parametric studies of the effect of MOx environment and control rods for PWR-UOx burnup credit implementation

    The increase of PWR-UOX fuel initial enrichment and the extensive needs for spent fuel storage or cask capacities reinforce the interest in taking burnup credit into account in criticality calculations. However, this utilization of credit for fuel burnup requires the definition of a methodology that ensures the conservatism of calculations. In order to guarantee the conservatism of the spent fuel inventory calculation, a depletion calculation scheme for burnup credit is under development. This paper presents the studies on the main parameters which have an effect on nuclides concentration: the presence of control rods during depletion and the fuel assembly environment, particularly the presence of MOx fuels around the UO2 assembly. Reactivity effects which are relevant to these parameters are then presented, and physics phenomena are identified. (author)

  14. Control of the neutronic and thermohydraulic conditions of power ramps in an irradiation loop for PWR fuel rod

    In order to study the power transients effects on PWR fuel rod clad, ramp tests in a pressurised water loop, are carried out at OSIRIS reactor. The present thesis deals with the on-line control of the device, during power ramp and conditioning irradiation. Based on a convolution-type resolution of the kinetics equations, a dynamic compensation of the Silver self-powered neutron detector was developed. With this method, the uncertainty of the ramp end-point is lower than 1%, thus it is very suited for monitoring both transient, as well as steady state conditions. Furthermore, a thermohydraulic model of the irradiation device is described: heat transfer equations, including gamma heating in materials, are solved to obtain temperatures and thermal fluxes of steady states. Results from the model and temperature measurements of the coolant are used together for fuel power determination, in real time. The clad external temperature profile also calculated and displayed, to improve the irradiation monitoring

  15. Porous Media Approach of a CFD Code to Analyze a PWR Component with Tube or Rod Bundles

    This paper presents a strategy to innovate CFD code into a PWR component analysis code. A porous media approach is adapted to two-fluid model and conductor model, and a pack of constitutive relations to close the numerical model into component analysis code. The separate verification calculations on open media, conductor model and porous media approach are introduced. Based on the CUPID code, the component analysis code has been developed. For porous media model, constitutive correlations of a two-phase flow regime map, interfacial area, interfacial heat and mass transfer, interfacial drag, wall friction, wall heat transfer and heat partitioning in flows through tube or rod bundles are added. Separate calculations were also conducted to verify the developed code

  16. PWR-UO2 nuclear fuel criticality study: control rod effects on infinite neutron multiplication factor and spent fuel composition

    Highlights: • A three-dimensional model of a PWR fuel were simulated. • Results using TRITON/T6-DEPL module in SCALE 6.0 and two libraries (238 and 44 groups) were compared. • Variations in the infinite neutron multiplication factor and the nuclides concentrations, both under control rod insertion effects were analysed. • Results show very good agreement with those published by OECD. -- Abstract: Deterministic and stochastic nuclear codes are software packages used to perform reactor physics calculations, especially in PWRs, the most common type of nuclear reactor currently in operation. The NEA Expert Group on Burn-up Credit Criticality Safety has published a Benchmark with results obtained from simulations of PWR-UO2 nuclear fuel. The same simulations were performed at DEN/UFMG with SCALE 6.0, a modular nuclear system code developed by Oak Ridge National Laboratory using two different neutron energy libraries (238 and 44 groups). The results obtained using a three-dimensional model with the T6-DEPL sequence of the TRITON module in SCALE 6.0 for spent fuel inventory and infinite neutron multiplication factor calculations show very good agreement with those published by the OECD. The main goal of this work is to validate the methodology at DEN/UFMG for future use in simulations related to Angra I, II and III Nuclear Power Plants

  17. Study for on-line system to identify inadvertent control rod drops in PWR reactors using ex-core detector and thermocouple measures

    Souza, Thiago J.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail: tsouza@nuclear.ufrj.br, E-mail: canedo@lmp.ufrj.br, E-mail: alessandro@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2015-07-01

    Accidental control rod drops event in PWR reactors leads to an unsafe operating condition. It is important to quickly identify the rod to minimize undesirable effects in such a scenario. In this event, there is a distortion in the power distribution and temperature in the reactor core. The goal of this study is to develop an on-line model to identify the inadvertent control rod dropped in PWR reactor. The proposed model is based on physical correlations and pattern recognition of ex-core detector responses and thermocouples measures. The results of the study demonstrated the feasibility of an on-line system, contributing to safer operation conditions and preventing undesirable effects, as its shutdown. (author)

  18. Development of dynamic control rod reactivity measurement methodology and computer code system for PWR

    Zee, Sung Quun; Lee, Chung Chan; Song, Jae Seung [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-09-01

    In order to apply dynamic control rod reactivity measurement (DCRM) method to domestic nuclear power reactor, the methodology of EPRC, 'Dynamic Reactivity Measurement of Rod Worth', was reviewed. It was also reviewed that items should be improve in three-dimensional kinetics code MASTER, which was developed by Korea Atomic Energy Research Institute, for use in DCRM. The validity of DORT two-dimensional synthesis method to calculate excore detector weighting factor were benchmarked via Yonggwang Unit 3 three-dimensional TORT calculation. The consistency of MASTER static core calculation results using neutron cross sections generated by commercial design tools PHENIX/ANC and DIT/ROCS were also verified via rodded and unrodded radial power distributions and control rod worth comparisons. 14 refs., 28 figs., 3 tabs. (Author)

  19. Studying approximating method and numerical computation of heat transfer of a fuel rod in PWR

    Based on the differential form of the general heat conduction equation, the approximating expression for a nu clear fuel rod was derived through integral. The fuel rod has asymmetrical heat resource distribution. Bessel function distribution is in radial direction and Cosine function distribution is in axis direction. Also, using the model of the advanced pressure water reactor 600, and taking an iterative calculation between tangential and normal diffusion terms in every control cell, temperature distribution of the fuel rod was computed by the finite volume method (FVM) in the unstructured grids. Comparing the approximate solutions with the numerical results, there was a good agreement between them. On this condition, we derived the location and size of maximum temperature by analysis the temperature distribution and variation. All of these can provide a useful reference for the pressure water reactor thermal design and thermal protection of nuclear engineering. (authors)

  20. The thermo-mechanics of the PWR fuel rod; La thermomecanique du crayon de combustible REP

    Barral, J.C. [Electricite de France, EDF, 75 - Paris (France); Gautier, B.; Chaigne, G. [Electricite de France, Service Etudes et Projets Thermiques et Nucleaires, 75 - Paris (France)] [and others

    1999-03-29

    The fuel rod mechanics is of a great importance in the safety and performance of the reactors. In this domain a meeting has been organized by the SFEN the 18 march 1998 at Paris. With the participation of scientists from CEA, EDF and Framatome, the physics of the fuel rods was presented based on four main aspects. Two first papers dealt with the solicitations of the fuel rod in normal and accidental conditions. The physical phenomena under irradiation were then detailed in the four following talks. Three papers presented the simulation and the codes of the fuel-cladding interactions with the diabolo effect. The last paper was devoted to the experiment feedback and the research programs. (A.L.B.)

  1. Study of friction and wear conditions of control rod drive and control cluster in PWR

    A good wear resistance is needed for control rod drive, control cluster and fixed guiding systems in internal structure of reactor primary circuit. In an experimental programme life-size material is verified with pressurized water in a loop called SUPERBEC at Cadarache and a research programme is set to find pairs of materials to reduce wear of fixed and moving parts. This programme is divided in three phases: first examination of control rod drive and guide wear, then reproducing wear in a laboratory test and study of material pairs and finally testing of materials in the pressurized water loop

  2. Water-side oxide layer thickness measurement of the irradiated PWR fuel rod by ECT method

    It has been known that water-side corrosion of fuel rods in nuclear reactor is accompanied with the metallic loss of wall thickness and hydrogen pickup in the fuel cladding tube. The fuel clad corrosion is one of the major factors to be controlled to maintain the fuel integrity during reactor operation. An oxide layer thickness measuring device equipped with ECT probe system was developed by KAERI, and whose performance test was carried out in NDT(Non-Destructive Test) hot-cell of PIE(Post Irradiation Examination) Facility. At first, the calibration/performance test was executed for the unirradiated standard specimen rod fabricated with several kinds of plastic thin films whose thickness were predetermined, and the result of which showed a good precision within 10% of discrepancy. And then, hot test was performed for the irradiated fuel rod selectively extracted from J44 fuel assembly discharged from Kori Unit-2. The data obtained with this device were compared with the metallographic results obtained from destructive examination in PIEF hot-cell on the same fuel rod to verify the validity of the measurement data. (author)

  3. Heat transfer in helium-cooled storage basket for spent PWR fuel rods

    The Environmental Technology grows nowadays to play an important role, so the storage technique for nuclear spent fuel is a key to supporting the reactor safety against the environmentalists. This paper addresses the thermal conduction in solid fuel, the thermal natural convection due to buoyancy and thermal radiation by using the commercial Computational Fluid Dynamics solver, Fluent 6 in the helium-cooled storage basket for nuclear reactor spent fuel rods. This study finds out the maximum temperature of the rods under the steady state balancing the total heat transfer rate, in addition, how much the radiation affects the overall heat transfer rate is envisaged. The basket contains the total 15x15 element rod fuel bundle. The decay heat generated on rod surface is transferred through the inner basket plate wall. The surface-to-surface model is selected as thermal radiation model. The helium contained and sealed initially inside the canister tank is treated as Boussinesq approximation assuming incompressible fluids in an enclosure sealed completely. Therefore, the overall solution process is carried out under steady state solution. The emissivity is constant as 0.8 for the fuel rod and 0.3 for the inner plate and other wall according to verification and validation assessments in Pacific Northwest Laboratory. And the calculation results are compared with that of experimentation. The deviation temperature differences are turned out to be a limited temperature more or less. The temperature gradient in radial direction of the basket shows the good agreement with the experiment. Almost of all heat power generated from the bundle is transferred via buoyancy coupled with thermal radiative heat transfer, here the ratio of thermal radiation is dependent on the emissivity. Temperature on basket surface affects the bundle temperature in basket. The total heat transfer rate of the fuel bundles is shared as natural circulation and thermal radiation, the heat balance sheet

  4. Analysis of a rod withdrawal in a PWR core with the neutronic- thermalhydraulic coupled code RELAP/PARCS and RELAP/VALKIN

    The Reactor Ejection Accident (REA) belongs to the Reactor Initiated Accidents (RIA) category of accidents and it is part of the licensing basis accident analyses required for pressure water reactors (PWR). The REA at hot zero power (HZP) is characterized by a single rod ejection from a core position with a very low power level. The evolution consists basically of a continuous reactivity insertion. The main feature limiting the consequences of the accident in a PWR is the Doppler Effect. To check the performance of the coupled code RELAP5/PARCS2.5 and RELAP5/VALKIN a REA in Trillo NPP is simulated. These analyses will allow knowing more accurately the PWR real plant phenomenology in the RIA most limiting conditions. (authors)

  5. New dynamic method to measure rod worths in zero power physics test at PWR startup

    To measure and validate the worth of control (or shutdown) bank in zero power physics test at PWRs, a dynamic control rod reactivity measurement (DCRM) technique has been developed and applied to six startups of Westinghouse plants as well as Korea Standard Nuclear power Plants based on the Combustion Engineering System 80 NSSS. With this technique, just one test bank is inserted into the bottom of the core at maximum stepping rate and withdrawn immediately to the all rod-out position. Specially designed inverse point kinetics equations are used to determine the test bank worth from the measured ex-core detector signals, which are controlled by the neutron-to-response conversion factor and the dynamic-to-static conversion factor. These two parameters are predetermined by the three-dimensional neutron adjoint flux distribution for both the top and bottom ex-core detector and the three-dimensional steady and transient core power distribution for test bank movement. To eliminate the gamma-ray effect on ex-core detector signals, a simple method, using reactivity curve characteristics, was also developed. To verify the DCRM method, a total of 28 bank worths of six different PWRs was measured by the DCRM and compared with those of conventional method. Results show that the DCRM method has a similar accuracy as the conventional technique. However, with the DCRM method, it only takes ∼15 min per bank from the beginning of rod insertion to the determination of measured static worth. From its performance, one can conclude that the DCRM method is an effective replacement for the conventional rod worth measurement method

  6. New dynamic method to measure rod worths in zero power physics test at PWR startup

    Lee, E.K. [Korea Electric Power Research Institute, 103-16 Munji Yusung, Daejeon 305 380 (Korea, Republic of)]. E-mail: lek@kepri.re.kr; Shin, H.C. [Korea Electric Power Research Institute, 103-16 Munji Yusung, Daejeon 305 380 (Korea, Republic of); Bae, S.M. [Korea Electric Power Research Institute, 103-16 Munji Yusung, Daejeon 305 380 (Korea, Republic of); Lee, Y.K. [Korea Electric Power Research Institute, 103-16 Munji Yusung, Daejeon 305 380 (Korea, Republic of)

    2005-09-15

    To measure and validate the worth of control (or shutdown) bank in zero power physics test at PWRs, a dynamic control rod reactivity measurement (DCRM) technique has been developed and applied to six startups of Westinghouse plants as well as Korea Standard Nuclear power Plants based on the Combustion Engineering System 80 NSSS. With this technique, just one test bank is inserted into the bottom of the core at maximum stepping rate and withdrawn immediately to the all rod-out position. Specially designed inverse point kinetics equations are used to determine the test bank worth from the measured ex-core detector signals, which are controlled by the neutron-to-response conversion factor and the dynamic-to-static conversion factor. These two parameters are predetermined by the three-dimensional neutron adjoint flux distribution for both the top and bottom ex-core detector and the three-dimensional steady and transient core power distribution for test bank movement. To eliminate the gamma-ray effect on ex-core detector signals, a simple method, using reactivity curve characteristics, was also developed. To verify the DCRM method, a total of 28 bank worths of six different PWRs was measured by the DCRM and compared with those of conventional method. Results show that the DCRM method has a similar accuracy as the conventional technique. However, with the DCRM method, it only takes {approx}15 min per bank from the beginning of rod insertion to the determination of measured static worth. From its performance, one can conclude that the DCRM method is an effective replacement for the conventional rod worth measurement method.

  7. Analysis of a control rod ejection accident in a 900 MWe PWR recycling plutonium with a gray control mode

    This research thesis addresses the study of the control rod cluster ejection accident in a 900 MWe PWR recycling plutonium and operating in grey mode, a class-IV accident in the safety report, which results from the failure of the cluster mechanism pressure enclosure, and results in a quick introduction of a reactivity within the core, and then in a violent power transient during which fuel strength can be put into question again. Two aspects are thus notably addressed: plutonium recycling, and grey mode operation. The objective is to qualitatively and quantitatively assess the evolution of physical parameters during the accident in order to determine the most severe scenarios and to be able to assess the severity of consequences. The author first studies all possible scenarios by means of a 2D+1D+0D calculation scheme in order to determine the most penalizing ones. Then, he develops a precise calculation based on 3D steady calculations, neutron kinetics calculations and thermal kinetics calculations in order to study the previously retained scenarios

  8. The development of flow test technology for PWR fuel assembly

    KAERI has an extensive program to develope PWR fuel assembly. In relation to the program, development of flow test technology is needed to evaluate the thermal hydraulic compactibility and mechanical integrity of domestically fabricated nuclear fuels. A high-pressure and high-temperature flow test facility was designed to test domestically fabricated fuel assembly. The test section of the facility has capacity of a 6x6 full length PWR fuel assembly. A flow test rig was designed and installed at Cold Test Loop to carry out model experiments with 5x5 rod assembly under atmosphere pressure to get information about the characteristics of pressure loss of spacer grids and velocity distribution in the subchannels. LDV measuring technology was established using TSI's Laser Dopper Velocimeter 9100-3 System

  9. Total evaluation of in bundle void fraction measurement test of PWR fuel assembly

    Nuclear Power Engineering Corporation is performing the various proof or verification tests on safety and reliability of nuclear power plants under the sponsorship of the Ministry of International Trade and Industry. As one program of these Japanese national projects, an in bundle void fraction measurement test of a pressurized water reactor (PWR) fuel assembly was started in 1987 and finished at the end of 1994. The experiments were performed using the 5 x 5 square array rod bundle test sections. The rod bundle test section simulates the partial section and full length of a 17 x 17 type Japanese PWR fuel assembly. A distribution of subchannel averaged void fraction in a rod bundle test section was measured by the gamma-ray attenuation method using the stationary multi beam systems. The additional single channel test was performed to obtain the required information for the calibration of the rod bundle test data and the assessment of the void prediction method. Three test rod bundles were prepared to analyze an axial power distribution effect, an unheated rod effect, and a grid spacer effect. And, the obtained data were used for the assessment of the void prediction method relevant to the subchannel averaged void fraction of PWR fuel assemblies. This paper describes the outline of the experiments, the evaluation of the experimental data and the assessment of void prediction method

  10. PWR rod ejection accident: uncertainty analysis on a high burn-up core configuration

    Le Pallec, J.C.; Studer, E.; Royer, E. [CEA Saclay, Direction de l' Energie Nucleaire, Service d' Etudes de Reacteurs et de Modelisation Avancee (DEN/SERMA), 91 - Gif sur Yvette (France)

    2003-07-01

    With the increasing of the discharge burn-up assembly, the rod ejection accident (REA) methodology based on the analyse of the hot spot from a decoupling methods of calculation does not allow to ensure the respect of safety criteria. The main reason is that the irradiated fuel certainly less solicited thermally is in the other hand more sensitive to a transient due to a rod ejection. Thus, the hot spot is not necessarily the sensitive point of the core. In the framework of high burn-up configurations, a new methodology tends to replace the former. It characterizes by the use of a best-estimate 3-dimensional modelling: coupling of the thermal hydraulics and neutronics, taking in account fuel properties depending on irradiation. To ensure the conservatism of the modelling response, this new approach has to be followed by an uncertainties analysis. Inputs from the benchmark RIA TMI-1 conducted by IRSN (France), NRC (United State of America) and KI (Russian) are used to perform a first analysis. The response of the modelling is the enthalpy deposited in an assembly. The analysis is based on the Design of Experiments (DoE) that permits to measure the weight of the main parameters and their interactions on the response. These last cannot be disregarded because they represent up to 20% of the penalizing uncertainty. This study shows that the main fuel modifications due to irradiation (radial power distribution, thermal properties degradation) have to be taken into account in a realistic thermal modelling during a strong transient.

  11. Assessment of a non-uniform heat flux correction model to predicting CHF in PWR rod bundles

    author for the prediction of CHF in a boiling channel with nonuniform axial heat flux distributions. In this study, we assess the applicability of the proposed model for PWR rod bundles. (authors)

  12. The development of flow test technology for PWR fuel assemblies

    The objective of this project is to design and construct a high temperature and pressure flow test facility and to develop flow test technology for the evaluation of PWR fuel performance. For the nuclear fuel safety aspect it is of importance to evaluate the thermalhydraulic compatibility and mechanical integrity of a newly designed fuel through the design verification test. The PWR-Hot Test Loop facility is under construction to be used to perform a pressure drop test, a lift force test and a fretting corrosion test of a fullsize PWR fuel assembly at reactor operating conditions. This facility was designed to be used to produce the hydraulic parameters of the existing PWR fuel assemblies(14x14FA, 16x16FA, 17x17FA) and to verify a design of advanced fuel assemblies (KAFA-I and KAFA-II) developed by KAERI. The PWR-Cold Test Loop facility with the 5x5 Rod Bundles in the test section was designed and installed to carry out the flow distribution study by means of Laser Doppler Velocimeter. The LDV techniques have been developed and used to measure the flow velocity and turbulent intensity for evaluating mixing effects of a newly designed spacer grid with and without mixing vanes, cross flow between the fuel assemblies and a turbulent model. (Author)

  13. Effect of boiling on the cladding corrosion of PWR fuel rod surface

    The demanding operational conditions in modern power plants (pressurized water reactors) can induce local boiling regimes at fuel rod surface in the hottest channels of the core (higher heat generation rate and primary coolant temperature). These new requirements for PWRs operating conditions may lead to accelerated corrosion kinetics of the Zircaloy-4 fuel cladding. The purpose of this thesis is to study the effect of boiling on the Zircaloy-4 cladding corrosion from tests performed in out-of-pile loops and conducted in severe chemical and thermohydraulic conditions (boiling, higher lithium content compared to PWRs...). The experimental results indicate an increase in external cladding corrosion under boiling conditions when lithium is present in the primary coolant. The higher is the lithium content in the coolant, the higher is the corrosion kinetics of the fuel cladding. Chemical analyses using Secondary Ion Mass Spectroscopy of zirconia films formed during these tests show that boiling leads to an enrichment of the chemical additives in the primary water (boron, lithium) at the surface of the cladding. It is demonstrated that this enrichment process is at the origin of an increase in the lithium incorporated in the oxide layers. Based on these results, the modelling of the chemical additives enrichment under boiling conditions is developed that allows to extend the COCHISE oxidation model to the prediction of Zircaloy-4 corrosion rates under two-phase flow heat transfer conditions. (author)

  14. Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit

    The Interim Staff Guidance on burnup credit (ISG-8) issued by the United States (U.S.) Nuclear Regulatory Commission's (NRC) Spent Fuel Project Office recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers. This recommended restriction eliminates a large portion of the currently discharged spent fuel assemblies from cask loading, and thus severely limits the practical usefulness of burnup credit. In the absence of readily available information on burnable poison rod (BPR) design specifications and usage in U.S. pressurized water reactors (PWRs), and the subsequent reactivity effect of BPR exposure on discharged spent nuclear fuel (SNF), NRC staff has indicated a need for additional information in these areas. In response, this report presents a parametric study of the effect of BPR exposure on the reactivity of SNF for various BPR designs, fuel enrichments, and exposure conditions, and documents BPR design specifications. Trends in the reactivity effects of BPRs are established with infinite pin-cell and assembly array calculations with the SCALE and HELIOS code packages, respectively. Subsequently, the reactivity effects of BPRs for typical initial enrichment and burnup combinations are quantified based on three-dimensional (3-D) KENO V.a Monte Carlo calculations with a realistic rail-type cask designed for burnup credit. The calculations demonstrate that the positive reactivity effect due to BPR exposure increases nearly linearly with burnup and is dependent on the number, poison loading, and design of the BPRs and the initial fuel enrichment. Expected typical reactivity increases, based on one-cycle BPR exposure, were found to be less than 1% Δk. Based on the presented analysis, guidance is offered on an appropriate approach for calculating bounding SNF isotopic data for assemblies exposed to BPRs. Although the analyses do not address the issue of validation of depletion methods for assembly designs with BPRs, they

  15. Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit

    Wagner, J.C.

    2001-09-28

    The Interim Staff Guidance on burnup credit (ISG-8) issued by the United States Nuclear Regulatory Commission's (U.S. NRC) Spent Fuel Project Office recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers. This recommended restriction eliminates a large portion of the currently discharged spent fuel assemblies from cask loading, and thus severely limits the practical usefulness of burnup credit. In the absence of readily available information on burnable poison rod (BPR) design specifications and usage in U.S. pressurized-water-reactors (PWRs), and the subsequent reactivity effect of BPR exposure on discharged spent nuclear fuel (SNF), NRC staff has indicated a need for additional information in these areas. In response, this report presents a parametric study of the effect of BPR exposure on the reactivity of SNF for various BPR designs, fuel enrichments, and exposure conditions, and documents BPR design specifications. Trends in the reactivity effects of BPRs are established with infinite pin-cell and assembly array calculations with the SCALE and HELIOS code packages, respectively. Subsequently, the reactivity effects of BPRs for typical initial enrichment and burnup combinations are quantified based on three-dimensional (3-D) KENO V.a Monte Carlo calculations with a realistic rail-type cask designed for burnup credit. The calculations demonstrate that the positive reactivity effect due to BPR exposure increases nearly linearly with burnup and is dependent on the number, poison loading, and design of the BPRs and the initial fuel enrichment. Expected typical reactivity increases, based on one-cycle BPR exposure, were found to be less than 1% {Delta}k. Based on the presented analysis, guidance is offered on an appropriate approach for calculating bounding SNF isotopic data for assemblies exposed to BPRs. Although the analyses do not address the issue of validation of depletion methods for assembly designs with BPRs

  16. Calibration of the Naval Postgraduate School 3.5 x 5.0 academic wind tunnel

    Nestor, Duane E.

    1990-01-01

    Approved for public release; distribution unlimited. The purpose of this thesis was to revitalize the Naval Postgraduate School's 3.5' x 5.0' academic wind tunnel. The wind tunnel had sustained previous damage to one of two sets of counter-rotating blades. Because of this change in configuration, a wind tunnel calibration was deemed necessary. Along with the calibration a digital data acquisition system was designed and implemented to aid in the data collection, storage and analysis for th...

  17. Pre-test prediction and post-test analysis of PWR fuel rod ballooning in the MT-3 in-pile LOCA simulation experiment in the NRU reactor

    The USNRC and the UKAEA have jointly funded a series of in-pile LOCA simulation experiments in the Canadian NRU reactor in order to secure further information on the thermal hydraulic and clad deformation response of PWR fuel rod bundles. Test MT-3 in the series was performed using reflood rate and rod internal pressure conditions specified by the UK nuclear industry. The parameters were selected to ensure the development of a near-isothermal clad temperature history during which zircaloy was required to balloon and rupture near the alpha-alpha/beta phase transition. Specification of the reflood rate conditions was assisted by the performance of a precursor test on an unpressurised rod bundle and by complementary application of appropriate thermal hydraulic analyses. Identification of the rod internal pressure needed to cause ballooning and rupture was achieved using a creep deformation model, BALLOON, in conjunction with the clad thermal history defined by the prior thermal hydraulic test. This paper presents the basis of the BALLOON analysis and describes its application in calculating the fill gas pressure for rods MT-3, their axial ballooning profile and the clad temperature at peak radial strain elevations. (author)

  18. 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

    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)

  19. Multi-group SP3 approximation for simulation of a three-dimensional PWR rod ejection accident

    Highlights: • The multi-group SP3 method developed and implemented in PARCS for the MOX analysis. • The verifications were performed in 2D and 3D, 2G and MG, diffusion and transport, with and without feedback. • All results show consistency with the reference results obtained from the ANL PN transport code VARIANT for steady-state and transport calculations. • It was found that the SP3 angular approximation captures sufficient transport effects for both steady-state and transient, and provides essentially the same results as the VARIANT P5 method. • From the transient results of the full-core problem, it was noted that MG is more conservative than 2G, and P1 is more conservative than SP3. - Abstract: Previous researchers have shown that the simplified P3 (SP3) approximation is capable of providing sufficiently high accuracy for both static and transient simulations for reactor core analysis with considerably less computational expense than higher order transport methods such as the discrete ordinate or the full spherical harmonics methods. The objective of this paper is to provide a consistent comparison of two-group (2G) and multi-group (MG) diffusion and SP3 transport for rod ejection accident (REA) in a practical light water reactor (LWR) problem. The analysis is performed on two numerical benchmarks, a 3 × 3 assembly mini-core and a full pressurized water reactor (PWR) core. The calculations were performed using pin homogenized and assembly homogenized cross sections for a series of benchmarks of increasing difficulty, in two-dimensional (2D) and three-dimensional (3D), 2G and MG, diffusion and transport, as well as with and without feedback. All results show consistency with the reference results obtained from higher-order methods. It is demonstrated that the analyzed problems show small group-homogenization effects, but relatively significant transport effects which are satisfactorily addressed by the SP3 transport method. The sensitivity tests

  20. Reactivity and neutron emission measurements of burnt PWR fuel rod samples in LWR-PROTEUS phase II

    Measurements have been made of the reactivity effects and the neutron emission rates of uranium oxide and mixed oxide burnt fuel samples having a wide range of burnup values and coming from a Pressurised Water Reactor (PWR). The reactivity measurements have been made in a PWR lattice moderated in turn with: water, a water and heavy water mixture, and water containing boron. An interesting relationship has been found between the neutron emission rate and the measured reactivity. (authors)

  1. PWR Fuel licensing in France - from design to reprocessing: licensing of nuclear PWR fuel rod design to satisfy with criteria for normal and abnormal fuel operation in France

    In this lecture are presented: French regulatory context; Current fuel management methods; Request from the french operator EdF; Most recent actions of the french Nuclear safety authority; Fuel assemblies deformations (impact of high burn-up; investigations during reactor's exploitation; control rods drop off times)

  2. Monitoring of garbage with a 5x5NaI (Tl) detector

    So far in that is carried out the first reload of nuclear fuel in the LVC, the monitoring of garbage has been carried out using monitors trade mark Eberline model RM 14. The procedure consists in manually monitoring each object and to separate of the considered 'clean' garbage the objects considered as contaminated, which register greater or equal counts to 100 cpm. This way to process was adequate under normal operation conditions, but not in the operation rhythm that implies a bigger maintenance since the time required for monitoring from 5 to 10 kg. of garbage is of the order of 0.5 hours and the production rhythm of this it ends up being a lot but high. Due to this necessity it was thought about the problem of looking by a more efficient monitoring method. In this work a method that uses a detector of NaI (Tl) of 5 x 5 inches is discussed. (Author)

  3. Development of CFD analysis to estimate flow behavior of PWR grid spacer

    PWR grid spacer has mixing vanes to promote coolant mixing around fuel rods. It is important to estimate flow field at downstream of the grid more precisely because the pressure loss and departure from nucleate boiling (DNB) of the grid are influenced by cross-flow and axial flow generated by mixing vanes. Mitsubishi has developed the Computational Fluid Dynamics (CFD) evaluation method for thermal-hydraulic design improvement of PWR grid spacer. The rod type Laser Doppler Velocimetry (rod LDV) is applied for cross-flow and axial flow measurements at rod gaps and sub-channels to develop the CFD flow field estimation at the downstream of the grid. Test results were compared to the data of Particle Imaging Velocimetry (PIV) technique, which were conducted by Westinghouse. Similarity of both measurements was confirmed qualitatively and some of data were in good agreement quantitatively. Comparison of measured data at the downstream of the grid and estimated values of CFD analyses, which were simulated LDV test conditions, were carried out. As the first step, parametric CFD analyses, which focused on spatial discretisation schemes, were performed by 2-sub channel model. At the second step, to simulate the test conditions and to compare quantitatively CFD and measurement for whole measured positions, flow field around 5x5 grid spacer and 5x5 rod bundle in the test section was modeled and the analysis was performed. It is confirmed that good agreement of CFD and measurements in most of positions but in several positions relatively larger deviation is observed. The causes of large deviation are discussed and future studies are planned. Velocity measurement program was conduced in collaboration with Westinghouse Electric Company. (author)

  4. Assessment of the uncertainties of COBRA sub-channel calculations by using a PWR type rod bundle and the OECD NEA UAM and the PSBT benchmarks data

    The assessment of the uncertainties of COBRA-IIIC thermal-hydraulic analyses of rod bundles is performed for a 5-by-5 bundle representing a PWR fuel assembly. In the first part of the paper the modeling uncertainties are evaluated in the term of the uncertainty of the turbulent mixing factor using the OECD NEA/NRC PSBT benchmark data. After that the uncertainties of the COBRA calculations are discussed performing Monte-Carlo type statistical analyses taking into account the modeling uncertainties and other uncertainties prescribed in the OECD NEA UAM benchmark specification. Both steady-state and transient cases are investigated. The target quantities are the uncertainties of the void distribution, the moderator density, the moderator temperature and the DNBR. We will see that - beyond the uncertainties of the geometry and the boundary conditions - it is very important to take into account the modeling uncertainties in case of bundle or sub-channel thermo-hydraulic calculations.

  5. Contact 1 and 2 experiments: behaviour of PWR fuel rod up to 15000 MWd.t-1

    In order to study the behaviour of 17 x 17 fuel rods, CEA and FRAMATOME have undertaken cooperative experiments. Some of them are called CONTACT 1 and CONTACT 2. In both experiments, a fuel rod with Zircaloy 4 cladding is irradiated in a pressurized water loop under 13 MPa. A power level 40 KW.m-1 for CONTACT 1 and 25 KW.m-1 for CONTACT 2 has been required. Each fuel rod is equipped with: a thermocouple in the center of the fuel, three pipes for the fission gases collection and measurement of pressure drop through the fuel stack, and a device with strain gauges in order to measure the evolution of the external cladding diameter. Results concern the centerline temperature evolution, the fuel to cladding gap variations and the kinetics of stable and radioactive fission gas release

  6. Temperature escalation in PWR fuel rod simulator bundles due to the zircaloy/steam reaction: Test ESBU-1

    This report describes the test conduct and results of the bundle test ESBU-1. The test objective was the investigation of temperature escalation of zircaloy clad fuel rods. The investigation of the temperature escalation is part of a program of out-of-pile experiments, performed within the framework of the PNS Several Fuel Damage Program. The bundle was composed of a 3x3 array of fuel rod simulators surrounded by a zircaloy shroud which was insulated with a ZrO2 fiber ceramic wrap. The fuel rod simulators comprised a tungsten heater, UO2 annular pellets, and zircaloy cladding over a 0.4 m heated length. A steam flow of 1 g/s was inlet to the bundle. The most pronounced temperature escalation was found on the central rod. The initial heatup rate of 20C/s at 11000C increased to approximately 60C/s. The maximum temperature reached was 22500C. The following fast temperature decrease was caused by runoff of molten zircaloy. Molten zircaloy swept down the thin cladding oxide layer formed during heatup. The melt dissolved the surface of the UO2 pellets and refroze as a coherent lump in the lower part of the bundle. The remaining pellets fragmented during cooldown and formed a powdery layer on the refrozen lump. The lump was sectioned posttest at several elevations: Dissolution of UO2 by the molten zircaloy, interaction between the melt and previously oxidized zircaloy, and oxidation of the melt had occurred. (orig.)

  7. Experimental investigation of the enthalpy- and mass flow-distribution in 16-rod clusters with BWR-PWR-geometries and conditions

    The enthalpy- and mass-flow-distribution at the outlet of two different 16-rod cluster test sections with uniform heating in axial and radial direction under steady state conditions has been measured for the first time by simultaneous sampling of 5 from 6 present characteristic subchannels in the bundle using the isokinetic technique and analysing the outlet quantities by a calorimetic method. The test-sections are provided with typical geometrical configurations for BWR s (70 bars; test section PELCO-S) and PWR s (160 bars; test-section EUROP). The latter has also been tested under BWR conditions (70 bars) to study the influence of geometry and pressure. The results showed the abnormal behaviour of the corner subchannel under BWR typical conditions (70 bars) which could not be found for PWR conditions (160 bars) and which is only an effect of pressure and not of geometry. The analysis of the experimental data confirms the usefullness of the subchannel sampling technique for the better understanding of the complex thermohydraulic phenomena under two-phase flow conditions in multirod bundles. Calculations of subchannel resistance coefficients for both types of spacers under one-phase flow conditions have been made with a special sub-structure method which showed a rather high local value of the corner subchannel. With the local drag coefficents the total resistance of the spacer has been evaluated and agreed well with measured values under adiabatic conditions. The measured subchannel data permit a direct valuation and examination of respective computer codes in a fundamental manner which are, however, not subject of this report

  8. Effect of applied current on the formation of defect in PWR nuclear fuel rods in resistance pressure welding process

    The welding of zirconium alloy components is one of the most critical processes in the fabrication of nuclear fuel rods used in pressurized water reactors. For this, various welding processes, such as gas tungsten arc welding, electron beam welding, laser beam welding, and resistance pressure welding (RPW), are used around the world. In Korea, the RPW process is being used to fabricate nuclear fuel assembly fuel rods. This study investigated changes in the weldment shape owing to welding conditions such as welding current, welding force, and overlapping. The welding soundness of the weldment was evaluated by hydraulic burst test. The welding temperature of the weld zone was measured using a thermal infrared method. Discontinuous black spots in the weld line, regarded as a non-bonding defect, were confirmed as spots caused by the carbide precipitation of zirconium during welding. (author)

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

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

  10. PWR FLECHT SEASET 21-rod-bundle flow-blockage task: data and analysis report. NRC/EPRI/Westinghouse report No. 11, main report and appendices A-J

    Loftus, M.J.; Hochreiter, L.E.; Lee, N.; McGuire, M.F.; Wenzel, A.H.; Valkovic, M.M.

    1982-09-01

    This report presents data and limited analysis from the 21-Rod Bundle Flow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Separate Effects and Systems Effects Test Program (FLECHT SEASET). The tests consisted of forced and gravity reflooding tests utilizing electrical heater rods with a cosine axial power profile to simulate PWR nuclear core fuel rod arrays. Steam cooling and hydraulic characteristics tests were also conducted. These tests were utilized to determine effects of various flow blockage configurations (shapes and distributions) on reflooding behavior, to aid in development/assessment of computational models in predicting reflooding behavior of flow blockage configurations, and to screen flow blockage configurations for future 163-rod flow blockage bundle tests.

  11. PWR FLECHT SEASET 21-rod-bundle flow-blockage task: data and analysis report. NRC/EPRI/Westinghouse report No. 11, main report and appendices A-J

    This report presents data and limited analysis from the 21-Rod Bundle Flow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Separate Effects and Systems Effects Test Program (FLECHT SEASET). The tests consisted of forced and gravity reflooding tests utilizing electrical heater rods with a cosine axial power profile to simulate PWR nuclear core fuel rod arrays. Steam cooling and hydraulic characteristics tests were also conducted. These tests were utilized to determine effects of various flow blockage configurations (shapes and distributions) on reflooding behavior, to aid in development/assessment of computational models in predicting reflooding behavior of flow blockage configurations, and to screen flow blockage configurations for future 163-rod flow blockage bundle tests

  12. The Influence Of Fuel-To-Clad Gap Anti) UO2 Grain Size On Fission Gas Release In High Burn-Up PWR Design Fuel Rods; IFA-519.9

    The experiment described in this report was designed to test the effect of gap size, and hence fuel temperatures, and grain size on the fission gas release characteristics of PWR design UO2 fuel rods. Three rods with different combinations of gap and grain size irradiated during several loadings in IFA-429 at an average linear heat rate below 20 kW/m. At this low power, there was minimal fission gas release. The burn-up achieved at the end of this irradiation was 26-29 MWd/kgUO2. The rods were reinstrumented with bellows type pressure transducers so that fission gas release could be monitored during irradiation from measurements of the rod internal pressure. The rods were loaded into IFA-519.9, and irradiated at a higher power level,≅ 40 kW/m. Irradiation at high power continued to ≅ 90 MWd/kgUO2. In-pile pressure data obtained from two rods confirmed a substantial fission gas release in both rods. The data have been analyzed using a simple diffusion and re-solution based release model incorporated into the FUEL TEMP-2 code. The predictions are in excellent agreement with the in-pile data and suffice to separate the individual effects of the two parameters on the fission gas release characteristics. It is concluded that in the present experiment, the effect of different gap sizes dominates that due to differences in UO2 grain size. (author)

  13. Uncertainty and Sensitivity of Neutron Kinetic Parameters in the Dynamic Response of a PWR Rod Ejection Accident Coupled Simulation

    C. Mesado

    2012-01-01

    Full Text Available In nuclear safety analysis, it is very important to be able to simulate the different transients that can occur in a nuclear power plant with a very high accuracy. Although the best estimate codes can simulate the transients and provide realistic system responses, the use of nonexact models, together with assumptions and estimations, is a source of uncertainties which must be properly evaluated. This paper describes a Rod Ejection Accident (REA simulated using the coupled code RELAP5/PARCSv2.7 with a perturbation on the cross-sectional sets in order to determine the uncertainties in the macroscopic neutronic information. The procedure to perform the uncertainty and sensitivity (U&S analysis is a sampling-based method which is easy to implement and allows different procedures for the sensitivity analyses despite its high computational time. DAKOTA-Jaguar software package is the selected toolkit for the U&S analysis presented in this paper. The size of the sampling is determined by applying the Wilks’ formula for double tolerance limits with a 95% of uncertainty and with 95% of statistical confidence for the output variables. Each sample has a corresponding set of perturbations that will modify the cross-sectional sets used by PARCS. Finally, the intervals of tolerance of the output variables will be obtained by the use of nonparametric statistical methods.

  14. Experimental investigation of a representative PWR nuclear fuel assembly spacer grid

    Castro, Higor Fabiano Pereira de; Mesquita, Amir Zacarias; Navarro, Moyses A.; Mattos, Joao Roberto Loureiro de; Santos, Andre A. Campagnole dos, E-mail: higorfabiano@hotmail.com, E-mail: amir@cdtn.br, E-mail: moysesnavarro@yahoo.com.br, E-mail: jrmattos@cdtn.br, E-mail: aacs@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The spacer grids are important structures present in nuclear fuel assembly from Pressurized Water Reactors (PWR). They play an important structural role and also assist in heat removal through the assembly by promoting increased turbulence of the flow. However the presence of spacer grids in the fuel assembly causes a localized pressure drop. In this paper we present experimental results of the water flow velocity profiles for five heights from a spacer grid present in a 5 x 5 rod bundle. These velocity profiles were obtained using a LDV (Laser Doppler Velocimetry). The tests were conducted for Reynolds numbers ranging from 1.8 x 10{sup 4} to 5.4 x 10{sup 4}. This experimental research was carried out in thermo-hydraulics laboratory of Nuclear Technology Development Center - CDTN. This experimental research also assists in CFD - Computational Fluid Dynamics numerical analysis process which is also developed at CDTN. (author)

  15. Isokinetic sampling of boiling R12 in a rod bundle: Methodology and first results

    In order to validate the computer codes using subchannel analysis in the P.W.R core calculations, a joint program, involving the three major actors of nuclear engineering in France (CEA, EDF and FRAMATOME), has been developed. The aim is to obtain reliable experimental data on distribution of flowrate and energy in the subchannels of a rod bundle in the range of thermal hydraulic parameters covered by the normal operating conditions of a P.W.R core. This paper essentially deals with the design of the experimental devices which have been developed. Nevertheless, some first results are also given. The first part gives details about the experimental set-up. To simplify the design, the tests are performed with Freon which is a well-known fluid for simulation of P.W.R core flows. As for test section, it is a 5 x 5 rod bundle (rods with 9.5 m in external diameter and 3.65 m in heating length) with uniform heat flux. In the first step of the study, grids have no mixing vane. The second part describes the whole design and the sampling methodology i.e.: the design of the sampling probes, the way to ensure isokinetic sampling, the devices for measurement of flow rate and enthalpy of sampled fluid. We also discuss the way to perform the tests to ensure that we get reliable data. Finally, the last part of the paper gives the first results obtained with thermal hydraulic conditions for pressure and flowrate closed to the nominal P.W.R. ones but with high heat flux (closed to CHF). Mass and heat balance, i.e. a comparison between the sum of flowrate and energy of sampled flows and the total flowrate and heat power, show clearly the consistency of our data. (author)

  16. Automatic classification of unexploded ordnance applied to Spencer Range live site for 5x5 TEMTADS sensor

    Sigman, John B.; Barrowes, Benjamin E.; O'Neill, Kevin; Shubitidze, Fridon

    2013-06-01

    This paper details methods for automatic classification of Unexploded Ordnance (UXO) as applied to sensor data from the Spencer Range live site. The Spencer Range is a former military weapons range in Spencer, Tennessee. Electromagnetic Induction (EMI) sensing is carried out using the 5x5 Time-domain Electromagnetic Multi-sensor Towed Array Detection System (5x5 TEMTADS), which has 25 receivers and 25 co-located transmitters. Every transmitter is activated sequentially, each followed by measuring the magnetic field in all 25 receivers, from 100 microseconds to 25 milliseconds. From these data target extrinsic and intrinsic parameters are extracted using the Differential Evolution (DE) algorithm and the Ortho-Normalized Volume Magnetic Source (ONVMS) algorithms, respectively. Namely, the inversion provides x, y, and z locations and a time series of the total ONVMS principal eigenvalues, which are intrinsic properties of the objects. The eigenvalues are fit to a power-decay empirical model, the Pasion-Oldenburg model, providing 3 coefficients (k, b, and g) for each object. The objects are grouped geometrically into variably-sized clusters, in the k-b-g space, using clustering algorithms. Clusters matching a priori characteristics are identified as Targets of Interest (TOI), and larger clusters are automatically subclustered. Ground Truths (GT) at the center of each class are requested, and probability density functions are created for clusters that have centroid TOI using a Gaussian Mixture Model (GMM). The probability functions are applied to all remaining anomalies. All objects of UXO probability higher than a chosen threshold are placed in a ranked dig list. This prioritized list is scored and the results are demonstrated and analyzed.

  17. GeoTIFF of 5x5 m Relative Reflectivity for Salt River Bay, St. Croix, 2011, UTM 20N NAD83 (NCEI Accession 0131858)

    National Oceanic and Atmospheric Administration, Department of Commerce — This image represents a LiDAR (Light Detection & Ranging) 5x5 meter resolution relative seafloor reflectivity surface for an area surrounding the mouth of Salt...

  18. GeoTIFF of 5x5 m Relative Reflectivity for Salt River Bay, St. Croix, 2011, UTM 20N NAD83

    National Oceanic and Atmospheric Administration, Department of Commerce — This image represents a LiDAR (Light Detection & Ranging) 5x5 meter resolution bathymetric surface for an area surrounding the mouth of Salt River Bay (SARI)St....

  19. GeoTIFF of 5x5 m Relative Reflectivity for Salt River Bay, St. Croix, 2011, UTM 20N NAD83

    National Oceanic and Atmospheric Administration, Department of Commerce — This image represents a LiDAR (Light Detection & Ranging) 5x5 meter resolution relative seafloor reflectivity surface for an area surrounding the mouth of Salt...

  20. GeoTIFF of 5x5 m Relative Reflectivity for Salt River Bay, St. Croix, 2011, UTM 20N NAD83 (NCEI Accession 0131858)

    National Oceanic and Atmospheric Administration, Department of Commerce — This image represents a LiDAR (Light Detection & Ranging) 5x5 meter resolution bathymetric surface for an area surrounding the mouth of Salt River Bay (SARI)St....

  1. SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

    Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

    1978-02-01

    The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocity and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.

  2. F.E.M. of PWR`s control rod cluster. Parametrical study of vibrating behavior by an Experiment Design method

    Bosselut, D. [Electricite de France (EDF), 92 - Clamart (France). Direction des Etudes et Recherches; Regnier, G. [Ecole Nationale Superieure des Arts et Metiers, 75 - Paris (France); Soulier, B. [DER Mecanique Pole Universitaire Leonard de Vinci, 92 - Paris (France)

    1997-03-01

    Some finite element models have been performed at EDF to simulate the vibrations of rod cluster and to analyse the wear phenomenon of rods using parametrical studies. In the first part, one of the finite element models is presented. The location of excitation sources is described. The calculated values are: rod displacement in the guiding cards, shock forces on the guiding cards and wear power produced. In the second part, a parametrical study is presented for a given computer experiment domain with an Experimental Design method. The building of the computer experiment design is described. The used polynomial model has all linear, quadratic and interactive terms for each of the 6 parameters (26 coefficients), 34 polynomials have been built to approach the effective shock forces and the mean wear power at each of the 17 guiding points. In the last part, the influence of parameters on calculated mean wear power is shown along rods and some responses surfaces are visualized. Systematism and closeness of experiment design technique is underlined. Easy simulation of all the response domain by polynomial approach, allows comparison with experiment feedback. (author) 9 refs.

  3. Hydraulic benchmark data for PWR mixing vane grid

    The purpose of the present study is to present new hydraulic benchmark data obtained for PWR rod bundles for the purpose of benchmarking Computational Fluid Dynamics (CFD) models of the rod bundle. The flow field in a PWR fuel assembly downstream of structural grids which have mixing vane grids attached is very complex due to the geometry of the subchannel and the high axial component of the velocity field relative to the secondary flows which are used to enhance the heat transfer performance of the rod bundle. Westinghouse has a CFD methodology to model PWR rod bundles that was developed with prior benchmark test data. As improvements in testing techniques have become available, further PWR rod bundle testing is being performed to obtain advanced data which has high spatial and temporal resolution. This paper presents the advanced testing and benchmark data that has been obtained by Westinghouse through collaboration with Texas A&M University. (author)

  4. Flaw detecting device for control rod cluster

    The device of the present invention measures the reduction of a wall thickness of each cladding tube of control rod caused by abrasion or flaws at a high accuracy in a short period of time in a control rod cluster for a PWR type reactor. Namely, a control rod cluster for a PWR type reactor is formed by bundling a plurality of control rods at one end. The reduction of the wall thickness of the control rod cladding tubes is measured by a flaw detecting method using eddy current. A group of flaw detecting probes is moved in the axial direction relative to the control rod cluster to scan all the control rods in the axial direction of the cladding tubes. The group of the flaw detecting probes has circular flaw detecting probes at least by the total number of the control rods corresponding to the position of each of the arranged control rods of the control rod cluster. Each of the circular flaw detecting probes has a measuring hole through which one control rod can pass. Accordingly, all of the control rods are scanned from one end to the other end for the control rod cluster thereby capable of measuring the entire surface simultaneously. (I.S.)

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

    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

  6. Modelling of pellet-cladding interaction in PWR's

    The pellet-cladding interaction that can occur in a PWR fuel rod design is modelled with the computer codes FRAPCON-1 and ANSYS. The fuel performance code FRAPCON-1 analyses the fuel rod irradiation behavior and generates the initial conditions for the localized fuel rod thermal and mechanical modelling in two and three-dimensional finite elements with ANSYS. In the mechanical modelling, a pellet fragment is placed in the fuel rod gap. Two types of fuel rod cladding materials are considered: Zircaloy and austenitic stainless steel. (author)

  7. Grid spacer effect on reflood behavior observed at reflood experiment with 5 x 5 bundle test section under wide pressure condition

    There are a lot of works on quenching behavior and minimum film boiling temperature. However, these phenomena are not clear enough for high-pressure condition. Therefore, reflood experiments were performed with 5 x 5-bundle test section under pressure of 2 MPa to 15.5 MPa. Geometry of a bundle test section used at the present experiments was principally the same as that of conventional PWRs. Grid spacers used at the present experiments were designed to simulate conventional ones. Data on clad temperature transient, quench time, and heat transfer coefficient were obtained. Quenching took place at several elevations where grid spacers located, and then each quenching front propagated downstream, or toward upper elevations under high mass flux condition. Thus, grid spacers showed influence of significant promotion of quenching occurrence above grid spacers. This promotion is more significant at higher mass flux and slightly significant at higher pressure. Under such condition, grid spacer effect is not negligible to analyze thermal-hydraulics accurately for PWRs and BWRs accidents. On the other hand, grid spacer effect on quenching occurrence is small at lower mass flux and lower pressure. The boundary of above two cases was determined experimentally. In order to study above characteristics, heat transfer coefficient distribution along the distance of the grid spacer was investigated. Heat transfer coefficient was enhanced above grid spacers and the enhancement was decayed along the distance of the grid spacer. The enhancement at the grid spacer was significant at higher mass flux. Relaxation length of grid spacer effect was about 0.1 m. Thus, effects of mass flux and pressure on promotion of quenching occurrence is explained by heat transfer enhancement above the grid spacers. (author)

  8. REWET, PWR LOCA accident experiments

    1 - Description of test facility: The REWET-II facility was designed for the investigation of the reflooding phase of a LOCA. The main design principle is the accurate simulation of the rod bundle geometry and the primary system elevations. This is necessary in order to have the correct flow channels and hydrostatic pressures for the reflooding process. The reactor vessel is simulated by a stainless steel U-tube construction consisting of downcomer, lower plenum, core and upper plenum. The primary loops contain a pipe simulating the broken loop and a connection line between the upper plenum and the downcomer simulating five intact loops. The containment is simulated by a pressure vessel (not in scale). Steam generators and primary pumps are simulated with flow resistances. The ECC-water can be injected to the downcomer and/or to the upper plenum by a pump or from an accumulator. All the elevation in the reactor vessel simulator are scaled to 1:1 (except the reactor upper head). The scale of the volumes and flow areas is 1:2333 referring to the number of the fuel rod simulators in the facility and the fuel rods in the reference reactor. The rod bundle is either in a hexagonal shroud, the inside distance of the opposite walls is 54.3 mm and the wall thickness 2 mm, or in a round shroud, the inner diameter is 66.0 mm and the wall thickness 2 mm. The simulation of the fuel-rod bundle consists of 19 indirectly electrically heated simulator rods. The heating coils are inside stainless steel claddings in magnesium oxide insulation. The heated length, the outer diameter and the lattice pitch of the fuel-rod simulators as well as the number (= 10) and construction of the rod bundle spacers are the same as in the reference reactor. The upper ends of the rods are attached to the upper tie plate. 2 - Description of test: Pressurized water reactors in use in Finland reactors have certain unique features which make them different from most other PWR designs. The 6 horizontal

  9. NSSR experiment with 50 MWd/kgU PWR fuel under an RIA condition

    Results obtained in the NSRR power burst experiments with irradiated PWR fuel rods with fuel burnup up to 50 MWd/kgU are described and discussed in this paper. Data concerning test method, test fuel rod, pulse irradiation, transient records during the pulse and post irradiation examination are described, and interpretations and discussions on fission gas release and fuel pellet fragmentation are presented. During the pulse-irradiation experiment with 50 MWd/kgU PWR fuel rod, the fuel rod failed at considerably low energy deposition level, and large amount of fission gas release and fragmentation of fuel pellets were observed. (author). 6 refs, 11 figs, 5 tabs

  10. Sucker rods

    Hoffmann, J.; Preis, L.

    1987-12-08

    The sucker rod system in a deep well sucker rod pump consists of a plurality of unidirectionally reinforced composite fiber rods extending substantially parallel but not in contact with each other, the cross-sectional area of which rods is less than 1 cm/sup 2/. This enables the advantageous material properties to be utilized to a high degree. The sucker rod system can be assembled on site. The individual composite fiber rods can be monitored when they are in the working position.

  11. Rodding Surgery

    ... a rod or nail into the internal cavity (medullary canal) of a long bone. Purpose of Rodding ... Osteogenesis Imperfecta: A Translational Approach to Brittle Bone Disease 1 st edition. New York, NY: Elsevier Academic ...

  12. DNB experiments for high-conversion PWR core design

    It is very important to clarify the departure from nucleate boiling (DNB) performance of core fuel assemblies for the high conversion pressurized water reactors (PWR). To investigate this, DNB experiments were performed in tight lattice rod bundles, using the model fluid Freon 12 and water under the actual operating conditions. In addition, DNB heat flux measurements in an annular-flow channel were carried out for the design of the fertile rods, which are installed in thimble tubes. (orig.)

  13. DNB experiments for high conversion PWR core design

    It is very important to clarify the departure from nucleate boiling (DNB) performance of core fuel assemblies for the high conversion PWR design. To investigate this, DNB experiments were performed in tight lattice rod bundles, using the model fluid freon-12 and the actual water. And also DNB heat flux mesurements in an annular flow channel were carried out for design of fertile rods which are installed in thimble tubes. (orig.)

  14. PWR and WWER thorium cycle calculation

    The first step of the investigation of the thorium fuel cycle with HELIOS 1.8 is validation of the results obtained from the code for this particular type of fuel. To complete this first task we performed calculation of the benchmark announced by IAEA in 1995. The benchmark was based on a simplified PWR model of the assembly with reduced fuel composition. This calculation was focused on a comparison of the methods and basic nuclear data. After successful validation of the code we focused our work on calculating the PWR and WWER thorium fuel cycles. The thorium cycle begins after the first use of UO2 fuel in the reactor as separation of plutonium from the burnt fuel. Separated plutonium is mixed with thorium and used as a new nuclear fuel in the reactor. For our calculation we prepared two variants of the assembly - the first variant is a homogeneous distribution and the second one is a non-homogenised distribution of thorium fuel in the assembly. The model of non-homogenised distribution of Pu-Th fuel was designed by replacing selected rods of the classical UO2 assembly by Pu-Th rods. These selected rods are distributed symmetrically in the assembly. Other rods in the assembly remain the same as in the classical UO2 assembly. The calculated and compared values are criticality and fuel composition as a function of burnup (Authors)

  15. Preliminary study of the economics of enriching PWR fuel with a fusion hybrid reactor

    This study is a comparison of the economics of enriching uranium oxide for pressurized water reactor (PWR) power plant fuel using a fusion hybrid reactor versus the present isotopic enrichment process. The conclusion is that privately owned hybrid fusion reactors, which simultaneously produce electrical power and enrich fuel, are competitive with the gaseous diffusion enrichment process if spent PWR fuel rods are reenriched without refabrication. Analysis of irradiation damage effects should be performed to determine if the fuel rod cladding can withstand the additional irradiation in the hybrid and second PWR power cycle. The cost competitiveness shown by this initial study clearly justifies further investigations

  16. Experience and reliability of Framatome ANP's PWR and BWR fuel

    Based on three decades of fuel supply to 169 PWR and BWR plants on four continents, Framatome ANP has a very large database from operating experience feedback. The performance of Framatome PWR and BWR fuel is discussed for the period 1992-2001 with special emphasis on fuel failures, countermeasures and their effectiveness. While PWR fuel performance in most reactors has been good, the performance in some years did suffer from special circumstances that caused grid-to-rod fretting failures in few PWRs. After solving this problem, fuel of all types showed high reliability again. Especially the current PWR fuel products AFA 3G, HTP, Mark B and Mark BW showed a very good operating performance. Fuel reliability of Framatome ANP BWR fuel has been excellent over the last decade with average annual fuel rod failure rates under 1x10-5 since 1991. More than 40% of all BWR fuel failures in the 1992-2001 decade were caused by debris fretting. The debris problem has been remedied with the FUELGUARDTM lower tie plate, and by reactor operators' efforts to control the sources of debris. PCI, the main failure mechanism in former periods, affected only 10 rods. All of these rods had non-liner cladding. (author)

  17. Thermohydraulic and constructional boundary conditions of an advanced PWR reactor

    The advantages and special features of an advanced PWR reactor (FDWR) have been systematically investigated for several years by the Department of Space Flight and Reactor Technology of the University of Brunswick (LRR-TUBS). The FDWR will have a homogeneous core, i.e. the fuel elements will consist of fuel rods of the same size and enrichment. (orig./GL)

  18. Radiological characterization of spent control rod assemblies

    Lepel, E.A.; Robertson, D.E.; Thomas, C.W.; Pratt, S.L.; Haggard, D.L. [Pacific Northwest Lab., Richland, WA (United States)

    1995-10-01

    This document represents the final report of an ongoing study to provide radiological characterizations, classifications, and assessments in support of the decommissioning of nuclear power stations. This report describes the results of non-destructive and laboratory radionuclide measurements, as well as waste classification assessments, of BWR and PWR spent control rod assemblies. The radionuclide inventories of these spent control rods were determined by three separate methodologies, including (1) direct assay techniques, (2) calculational techniques, and (3) by sampling and laboratory radiochemical analyses. For the BWR control rod blade (CRB) and PWR burnable poison rod assembly (BPRA), {sup 60}Co and {sup 63}Ni, present in the stainless steel cladding, were the most abundant neutron activation products. The most abundant radionuclide in the PWR rod cluster control assembly (RCCA) was {sup 108m}Ag (130 yr halflife) produced in the Ag-In-Cd alloy used as the neutron poison. This radionuclide will be the dominant contributor to the gamma dose rate for many hundreds of years. The results of the direct assay methods agree very well ({+-}10%) with the sampling/radiochemical measurements. The results of the calculational methods agreed fairly well with the empirical measurements for the BPRA, but often varied by a factor of 5 to 10 for the CRB and the RCCA assemblies. If concentration averaging and encapsulation, as allowed by 10CFR61.55, is performed, then each of the entire control assemblies would be classified as Class C low-level radioactive waste.

  19. 压水堆核电站棒位探测器样机设计及试验研究%Prototype Design and Experimental Study of Rod Position Detector of PWR Nuclear Power Station

    白冰; 周建明; 吕永红

    2013-01-01

    Start and stop , power regulation of nuclear reactors depends on the control rod drive mechanism driv -ing control rod up and down movements .Accurate and reliable measurement of control rod position is an impor-tant guarantee for the safe operation of the reactor .In this paper , in the process of designing and constructing the rod position detector prototype of nuclear power station , the measuring principle , coding , coil number of choices , coil frame are described , the system is tested and the test results show that , the prototype design is reasonable and it has reliable performance .%核电站反应堆的启停、功率调节依靠控制棒驱动机构驱动控制棒上下运动来实现,控制棒位置的准确可靠测量是反应堆安全运行的重要保证。论文借助二代加核电站棒位探测器工程样机的研制,详细介绍了在棒位探测器设计过程中测量原理、编码方式、线圈数量的选择以及线圈骨架结构等内容,并通过各项性能测试与试验结果的分析,得出样机设计合理与性能可靠的结论。

  20. Progress of PWR reactor fuels: OSIRIS equipments

    The experimental reactor Osiris situated at the Saclay Nuclear Centre is a reactor fitted with tests and monitoring facilities. Of the pool and open core type, it can test the test fuel of PWR power stations under high neutron flux. The characteristic stresses of the operating states of power reactors can be reproduced in experimental devices suited to the various study subjects, be this the creep and deformation of zircaloy claddings, the behavior of fuel rods to power ramps, to load following, to remote regulation, to the cooling state in double phase or just analytical tests. The experimental irradiation devices extend from the single static coolant capsule, such as the NaK alloy, to the dynamic coolant test loop that operates in the cooling conditions representative of PWR's including water chemistry. Ancillary devices make it possible to carry out examinations and non-destructive testing: immersed neutron radiography, gamma scanning visualization monitoring device, eddy currents, profilometering

  1. Vibrational characteristics and wear of fuel rods

    Fuel rod wear, due to vibration, is a continuing concern in the design of liquid-cooled reactors. In my report, the methodology and models that are used to predict fuel rod vibrational response and vibratory wear, in a light water reactor environment, are discussed. This methodology is being followed at present in the design of Westinghouse Nuclear Fuel. Fuel rod vibrations are expressed as the normal bending modes, and sources of rod vibration are examined with special emphasis on flow-induced mechanisms in the stable flow region. In a typical Westinghouse PWR fuel assembly design, each fuel rod is supported at multiple locations along the rod axis by a square-shaped 'grid cell'. For a fuel rod /grid support system, the development of small oscillatory motions, due to fluid flow at the rod/grid interface, results in material wear. A theoretical wear mode is developed using the Archard Theory of Adhesive Wear as the basis. Without question certainty, fretting wear becomes a serious problem if it progresses to the stage where the fuel cladding is penetrated and fuel is exposed to the coolant. Westinghouse fuel is designed to minimize fretting wear by limiting the relative motion between the fuel rod and its supports. The wear producing motion between the fuel rod and its supports occurs when the vibration amplitude exceeds the slippage threshold amplitude

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

    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

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

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

    2013-10-15

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

  4. Post DNB heat transfer experiments for PWR fuel assemblies

    Nuclear Power Engineering Corporation (NUPEC) and Mitsubishi performed heat transfer experiments on post DNB (departure from nucleate boiling) for the pressurized water reactor (PWR) fuel assemblies under the sponsorship of the Japanese Ministry of Economy, Trade and Industry (METI) as one of a series of fuel assembly verification tests. Based on the obtained experimental data, a new evaluation model for the fuel rod heat transfer behavior after DNB was developed. A large safety margin, which had remained in the present thermal-hydraulic design that did not allow DNB, was confirmed by applying the developed model to the PWR plant safety analysis. (author)

  5. Effect of startup ramp rate on PWR fuel reliability

    A wide range of startup strategies and restart times currently exists for commercially operated pressurized water reactors (PWRs). The variability in PWR restart strategies is a function of several factors, including reactor system instrument calibration, primary and secondary water chemistry control, and vendor specified fuel rod ramp rate limitations. Fuel vendors, as a means to mitigate pellet-cladding interaction (PCI) leading to fuel rod failures, specify reactor power ramp rate limitations following a refueling outage. Typical restart ramp rates range between 3% per hour and 4% per hour of full reactor power above a threshold reactor power level between 20% and 40% full power. This paper summarizes an analytical evaluation performed to assess the technical basis for PWR restart ramp rate restrictions and to provide the technical justification to propose less restrictive power ramp rate conditions. Two combinations of PWR reactor types (Yonggwang Unit 2 and 4) and fuel rod designs were used to evaluate the impact of ramp rate and threshold power conditions on the PCI behavior of once-burned and twice-burned fuel rods. The fuel rod condition at the reactor restart of interest was established using the ESCORE steady state fuel performance program. Detailed PCI calculations were performed using the FREY fuel rod behavior program. The assessment identified significant margin to PCI failure for current ramp rate conditions used in YGN Unit 2 and 4. Based on the analytical evaluation presented, ramp rates up to 5% per hour above threshold power levels up to 60% of full reactor power can be used without concern for fuel rod integrity during reactor restarts following a refueling outage

  6. Rod worth measurement innovation at Westinghouse

    Bank worth measurement of control rods and shut-down rods is required for every cycle startup of a nuclear power plant for design validation. For Pressurized Water Reactors (PWR), the bank worth measurement is part of the Low Power Physics Tests (LPPT) program. In almost all instances, this program is on critical path during ascension to power. There is a strong incentive for the utility industry to have a fast and reliable method of measuring the bank worth. Over the past decade, Westinghouse has been developing new advanced rod worth measurement methods to provide faster, safer, more accurate and easier to use products. The advancement of 3D core simulation codes has made it possible to make revolutionary developments for a new generation of rod worth measurement methods

  7. Fabrication of PWR fuel assembly and CANDU fuel bundle

    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

  8. Control rod

    In a control rod of a nuclear reactor, B-10 is distributed such that the concentration of B-10 in boron carbide is lowered from the outer side to the inner side of the control rod. Alternatively, the inside of a blade is radially divided into a plurality of regions, and the amount of boron carbide loaded in the regions is reduced from the outer side to the inner side of the control rod. Alternatively, a plurality of sintered products of boron carbide are disposed radially in the blade, and the sintered product is divided to a first region where the B-10 content is relatively low and a second region having a higher B-10 content than the first region, and the sintered product disposed on the inner side is constituted such that the position of the first region in the sintered product is localized to the inner side of the control rod. Then, improvement of reliability and reduction of cost can be attained while maintaining an effective neutron absorbing performance of control rods taking neutron flux distribution into consideration. (N.H.)

  9. RANS modeling for flow in nuclear fuel bundle in pressurized water reactors (PWR)

    This paper presents use of Reynolds-Averaged Navier-Stokes (RANS) based turbulence model for single-phase CFD analysis of flow in Pressurized Water Reactor (PWR) Assemblies. An open source code called OpenFoam was used for computational fluid dynamics (CFD) study using computational meshes generated using Shari Harpoon. The PWR assembly design used in this analysis represents a 5 x 5 pin design including structural grid equipped with mixing vanes. The design specifications used in this study were obtained from the experimental setup at Texas A&M University and the results obtained are used to validate the CFD software, algorithm, and the turbulence model used in this analysis. (author)

  10. In-pile test of Qinshan PWR fuel bundle

    In-pile test of Qinshan Nuclear Power Plant PWR fuel bundle has been conducted in HWRR HTHP Test loop at CIAE. The test fuel bundle was irradiated to an average burnup of 25000 Mwd/tU. The authors describe the structure of (3 x 3-2) test fuel bundle, structure of irradiation rig, fuel fabrication, irradiation conditions, power and fuel burnup. Some comments on the in-pile performance for fuel bundle, fuel rod and irradiation rig were made

  11. Analysis of dynamic behavior of a PWR utilizing the computer program SARDAN 2

    In the design of a PWR nuclear plant it is necessary to verify if the design limits are respected, even under abnormal operation condition. An evolution of SARDAN code, developed to simulate transients in PWR, are presented. The new aspects incorporeted in SARDAN 2 are: the fuel ROD analysis in finite-diference, an open channel model for the critic subchannel analysis and the introduction of a simplified model for the automatic control system. The program has been tested in accident condition II, in special, uncontrolled ROD cluster assembly bank withoraw, dropped full-length assembly group, uncontrolled Boron dilution, and the results obtained were considered satisfactory. (Author)

  12. Recent development for improving of PWR flexibility to load follow and frequency control operation

    In order to adjust the PWR electricity generation to the consumption network, new operating conditions were established. Those new conditions generate additional mechanical and thermal sollicitations due to the frequent motion of control rod banks, consisting of mechanical fatigue cycling and wear at the level of control rode drive mechanisms, control rods and guide tubes, wear and thermal fatigue cycling at the level of fuel assemblies. This paper presents the various aspects of this program including identification of the most critical areas of components, basic research in laboratories for resolving wear problems in PWR environment, improvement of local hydraulics for reducing loads, and endurance testing of full scale components on testing facilities

  13. Control rod drive mechanism test program. Revision 3

    A description is given of the testing and development of three control rod drive mechanisms for use on commercial PWR plants designed by B and W. The test results indicate that all three drives are reliable and ensure safe, dependable reactor operation

  14. SEFLEX fuel rod simulator effects in flooding experiments. Pt. 1

    The aim of the SEFLEX program has been to quantify the influence of the design and the physical properties of different fuel rod simulators on heat transfer and quench front progression in unblocked and blocked rod bundles during the reflood phase of a LOCA in a PWR. Fuel rod simulators with Zy claddings and a gas-filled gap between claddings and pellets exhibit lower peak cladding temperatures and shorter quench times than gapless heater rods with stainless steel claddings. Grid spacers cause significant cooling enhancement downstream during the time span at which maximum cladding temperatures occur. Ballooned Zy claddings, forming e.g. a 90 percent blockage, are quenched substantially earlier than thickwall stainless steel blockage sleeves attached to the rods, and even earlier than undeformed rod claddings. A comparison of test data with results of the 'Best Estimate' computer program COBRA-TF shows a good agreement with unblocked bundle data including grid spacer effects. (orig./HP)

  15. PREP-PWR-1.0: a WIMS-D/4 pre-processor code for the generation of data for PWR fuel assemblies

    The PREP-PWR-1.0 computer code is a substantially modified version of the PREWIM code which formed part of the original MARIA System (Report J.E.N. 543). PREP-PWR-1.0 is a comprehensive pre-processor code which generates input data for the WIMS-D/4.1 code (Report PEL 294) for PWR fuel assemblies, with or without control and burnable poison rods. This data is generated at various base and off-base conditions. The overall cross section generation methodology is described, followed by a brief overview of the model. Aspects of the base/off-base calculational scheme are outlined. Additional features of the code are described while the input data format of PREP-PWR-1.0 is listed. The sample problems and suggestions for further improvements to the code are also described. 2 figs., 2 tabs., 12 refs

  16. Post-DNB heat transfer experiments under PWR operating conditions in annular test sections

    Single channel post-DNB heat transfer tests were carried out in the high-pressure and high-flow region which bounds PWR operating conditions. They were planned as preliminary tests prior to testing rod bundles. The test section consists of a flow channel with an ID of 15.9mm and a heater rod with an OD of 10.7mm and a heated length of 1.5m. Two types of heater rods, using direct- and indirect- heating, were compared and the indirect sheathed heater has been chosen for the next rod bundle test. The test data were compared with several existing film boiling heat transfer correlations. It was shown that post-DNB heat transfer under PWR conditions is better than predicted by existing film boiling correlations. (author)

  17. PWR type reactor

    Coolant discharging windows disposed to a control rod cluster guide tube are distributed in a region between the height of the lower end of a coolant exit nozzle and the height of the lower nozzle of an upper reactor core support column. The flow of coolants in the lateral direction toward an exit nozzle does not flow backwardly from the discharging windows to the inside of the control rod cluster guide tube, and the flow of coolants in the control rod cluster guide tube is discharged from each of the coolant discharging windows to the outside directly and rapidly while forming branched streams. As a result, the flow rate of coolants passing through a continuous portion is greatly reduced, and the flow rate of coolants in the direction traversing the control rods is greatly reduced. Accordingly, fluid vibrations for all the control rod clusters is reduced to reduce abrasion and the thickness reduction of the walls of a guide plate of the control rod cluster guide tube caused by contact with the control rods. (N.H.)

  18. Non-destructive Testing Dummy Nuclear Fuel Rods by Neutron Radiography

    WEI; Guo-hai; HAN; Song-bai; HE; Lin-feng; WANG; Yu; WANG; Hong-li; LIU; Yun-tao; CHEN; Dong-feng

    2013-01-01

    As a unique non-destructive testing technique,neutron radiography can be used to measure nuclear fuel rods with radioactivity.The images of the dummy nuclear fuel rods were obtained at the CARR.Through imaging analysis methods,the structure defections,the hydrogen accumulation in the cladding and the 235U enrichment of the pellet were studied and analyzed.Experiences for non-destructive testing real PWR nuclear fuel rods by NR

  19. The integrated PWR

    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. Plutonium recycling in PWR

    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)

  1. Multi-rod burst test under a loss-of coolant accident condition, (4)

    Multi-rod burst test of No.7808 bundle was performed in steam to estimate quantitative coolant flow channel restriction caused by the ballooning of zircaloy claddings in a fuel assembly during a LOCA transient in LWRs. The test was conducted under the condition that the initial internal pressure in each rod was 35kg/cm2 (RT) and the heating rate was 90C/s in steam with flow rate of 0.4g/cm2.min. The following results were obtained; (1) Maximum and burst pressures in rods were in the range 45 to 48kg/cm2 and 41 to 45kg/cm2, respectively. The burst temperature of cladding were estimated to be 850 to 8800C. (2) Axial portions of tubes with greater than 34% strain were observed in the range 0 to 40mm in most rod. The mean length was 19mm in the bundle. (3) The degree of maximum increase in cross-sectional area is 54.2% in the bundle(7 x 7) and 66.9% in the internal rods(5 x 5). (4) Maximum channel area restriction was 40.5% in the bundle(7 x 7) and 51.4% in the internal rods(5 x 5). (author)

  2. Analysis of reactivity insertion accidents in PWR reactors

    A calculation model to analyze reactivity insertion accidents in a PWR reactor was developed. To analyze the nuclear power transient, the AIREK-III code was used, which simulates the conventional point-kinetic equations with six groups of delayed neutron precursors. Some modifications were made to generalize and to adapt the program to solve the proposed problems. A transient thermal analysis model was developed which simulates the heat transfer process in a cross section of a UO2 fuel rod with Zircalloy clad, a gap fullfilled with Helium gas and the correspondent coolant channel, using as input the nulcear power transient calculated by AIREK-III. The behavior of ANGRA-i reactor was analized during two types of accidents: - uncontrolled rod withdrawal from subcritical condition; - uncontrolled rod withdrawal at power. The results and conclusions obtained will be used in the license process of the Unit 1 of the Central Nuclear Almirante Alvaro Alberto. (Author)

  3. CHF and flow instability in rod bundles

    Data for two very different rod bundles have been analyzed using a new CHF correlation and a crude, but simple, subchannel analysis. The CHF correlation was developed for round uniform tubes and has been shown to accurately predict CHF in nonuniform tubes. The first set of data was for a KWU rod bundle (37 rods) with a heated length of 3.00 m and an O.D. (outside diameter) of 12.9 mm over a range of pressure 70 to 150 bar in upflow. The second set of data was for a 5 x 5 TRIGA rod bundle with a heated length of 0.559 m and 13.75 mm O.D. over a range of pressure of 0.945 to 1.372 bar in downflow. In contrast to the KWU data, the correlation greatly over estimates the CHF values for the TRIGA data. The TRIGA CHF data correlate very well with the variable qsat assuming no mixing, qc,exp = 0.955qsat (stdev = 9.87%). This result strongly suggests that these instabilities, which resulted immediately in CHF, are triggered by the Onset of Flow Instability (OFI) rather than CHF. The wide spread in rod power factors, the low pressure, and the downflow condition all contribute to promoting this type of instability (Ledinegg). The crude subchannel analysis has been compared with calculations of exit conditions of the hot channel using COBRA code. The agreement is fair when the homogeneous equilibrium model is used in the COBRA code. This is expected since the exit of the hot channel is always subcooled. Using Zuber's, along with other, void fraction relations in COBRA yields much lower exit velocities and high positive exit qualities, and, in some cases, convergence difficulties arise. The facts indicate that the bundle has already past the OFI point: which is possible since no CHF calculation was made in these COBRA analyses. (J.P.N)

  4. Fuel rod

    The present invention provide a fuel rod used in a BWR type reactor, preventing the occurrence of defects of weld portions and improving the operationability of test and assembling operation to improve the quality of weld portions. Namely, the fuel rod is formed by loading a plurality of fuel pellets in a cladding tube. The outer diameter of a groove portion of a tightly sealing end plug to be inserted and welded to the open end of the cladding tube is made substantially identical with the inner diameter of the cladding tube. A neck portion having a diameter smaller than the outer diameter of the groove portion is disposed between an end plug main body and the groove portion. As a result, since the outer diameter of the groove portion is substantially identical with the inner diameter of the cladding tube, the positioning is facilitated. Since the neck portion having a smaller diameter than the outer diameter of the groove portion is disposed in the groove portion, a gap is formed in the welded portion thereby enabling to facilitate the confirmation of weld sag for confirming integrity of the weld by a non-destructive test. (I.S.)

  5. Zircaloy-4 hydriding. Hydrogen distribution in PWR's rod cladding

    In pressurised water reactors, Zircaloy 4 is used as fuel cladding in contact with hot water. The precipitation of hydrides at room temperatures causes mechanical deterioration of the cladding. As the cladding is subjected to a radial temperature gradient, the hydrogen distribution is greatly affected. The image analysis method is used to determine the hydride distribution in the irradiated cladding. To calibrate this method, a device was specially built for the preparation of Zircaloy specimens with known hydrogen contents. The hydriding conditions and hydrogen content determination procedures were fixed. We have successfully realized specimens with various hydrogen contents. With these specimens, a relationship between the parameter Sv (surface density of hydrides) and the hydrogen content was established. This parameter Sv is independent from the Zircaloy 4 metallurgical state (i.e. stress relieved or recrystallized) and from the analysis section (longitudinal or transverse). Study of hydrogen content and hydride distribution in irradiated cladding by means of image analysis showed that the method is limited by its ability of separation between neighbouring hydrides at cladding's periphery where the hydrogen content can reach several thousands ppm. Nevertheless, this method gives us some information about hydride distribution inside the cladding. A model for thermal diffusion was developped to stimulate the migration of hydrogen in Zirconium alloys. This model was used to predict hydrogen distribution in the irradiated cladding. Comparison of model predictions with results of image analysis shows good agreement. (Author). refs., figs., tabs

  6. Modelling the waterside corrosion of PWR fuel rods

    The mechanism of zirconium alloy cladding corrosion in PWRs is briefly reviewed, and an engineering corrosion model is proposed. The basic model is intended to produce a best-estimate fit to circumferentially-average oxide thickness measurements obtained from inter-span positions, way from the effects of structural or flow mixing grids. The model comprises an initial pre-transition weight gain expression which follows cubic rate kinetics. On reaching a critical oxide thickness, a transition to linear rate kinetics occurs. The post-transition corrosion rate includes a term which is dependent on fast neutron flux, and an Arrhenius thermal corrosion rate which has been fitted to isothermal ex-reactor data. This thermal corrosion rate is enhanced by the presence of lithium in the coolant, and by the concentration of hydrogen in the cladding. Different cladding materials are accounted for in the selection of the model constants, and results for standard Zircaloy-4, low tin (or ''optimized'') Zircaloy-4, and the Westinghouse advanced alloy ZIRLOTM are presented. A method of accounting for the effects of grids is described, and the application of the model within the ENIGMA-B and ZROX codes is discussed. (author). 35 refs, 6 figs, 3 tabs

  7. Whole-rod testing of intact and defective LWR rods under expected dry-storage conditions

    The objective of this project is to provide the Nuclear Regulatory Commission with information to confirm or establish spent fuel dry storage licensing positions relative to: (1) the long-term, low-temperature (less than 2500C) behavior of spent fuel rods in dry storage; and (2) the radioactive contamination potential of crud from cladding in dry storage. The basic need for this data is to: confirm long-term, low temperature (less than 2500C) spent fuel dry storage performance predictions based on theoretical analyses and on results from high-temperature, short-term laboratory tests; determine the nature and behavior of crud layers as a function of dry storage time; and determine the potential radioactive crud contamination (e.g., spalling characteristics) for dry storage. An eight-rod test matrix of PWR and BWR rods was chosen which consisted of all combinations of intact or breached cladding in an oxidizing or inert atmosphere. The PWR rods (30.5 GWD/MTU) were discharged from H.B. Robinson in May 1974, and the BWR rods (12.9 GWD/MTU) were discharged from Peach Bottom in March 1976. The eight test rods were visually inspected for crud and defects with the results recorded on video tape. Cladding penetration was confirmed. All the rods were put in test capsules with the appropriate atmosphere and leak checked. The test capsules were loaded into a test train and the train was placed in the furnace cavity. The test was started on September 15, 1982 and is presently at 2300C. After the first 10-month run is completed, an interim examination, consisting of visual inspection, gamma scanning, and crud sampling, will be conducted

  8. Abrasion measuring method for rod of control rod assembly of reactor

    The present invention provides a method of easily measuring abrasion caused on the outer surface of control rods of a control assembly to be used in a PWR type reactor. Namely, the control rod assembly comprise a plurality of control rods assembled in a cluster-like manner. Light is irradiated to a control rod to be measured from an optical measuring device for measuring the extent of abrasion on the surface of the control rods. The distance is measured by receiving the reflected light. The depth of abrasion is determined by comparing the thus measured distance to the abraded portion and the distance to an integral portion. Then, the depth of the abrasion is adjusted based on the control rod position and the angle to determine final depth of abrasion. The abrasion of control rods can be measured by remote control using one kind of light sensor. The device can be reduced in the size and the time for the measuring operation can also be shortened. (I.S.)

  9. Enhancing heat transfer and crud mitigation in PWR fuel

    This paper discusses three methods for increasing single phase heat transfer in PWR fuel. The primary effect of increasing heat transfer is a reduction in the steaming rate from the fuel rods, which in turn reduces the likelihood of crud formation on the fuel rods and the potential for adsorption of boron into the crud. The advantage of lowering boron mass on the fuel is reduced risk of Axial Offset Anomaly (AOA). Another benefit of reduced crud formation is a lower risk of localized corrosion, a known contributor to rod cladding failures. Thinner crud leads to locally lower rod operating temperatures (lower corrosion rate) since crud acts as a thermal insulator between the rod and the coolant. The first method of increasing heat transfer involves addition of more than one Intermediate Flow Mixing vane grid (IFM) in the span between two neighboring structural spacing grids. The second method includes optimization of the mixing vane according to axial position. The third method involves variation of the IFMs axial position to optimize axial distribution of rod heat transfer. (authors)

  10. Sub Channel Thermal hydraulics Design Analysis of PWR-KSNP

    Sub channel analysis for the fuel element of thermal hydraulics design PWR-KSNP reactor has been carried out. PWR-KSNP reactor is a kind of Pressurized Water Reactor (PWR) Nuclear Power Plant developed by Korea (Korean Standard Nuclear Plant), that produce an electricity power about 1000 MWe. In the analysis, a fuel assembly with 4 fuel rods piled up into matrix 2 x 2, and surrounding by 9 sub channels of coolant, was used as a calculation model. There are 3 models of fuel assembly, i.e. the radial factors in the first model are 1.144, 1.144, 1.120 and 1.121, in the second fuel model are 0.994 , 1.005 , 0.987 and 0.989, and in the third model are 2.500, 1.144, 1.120 and 1.121, respectively. The calculated results using the COBRA IV-I code showed that the maximum cladding temperature revolved by 340.3 - 349.0 ℃, the maximum temperature of meat surface (outer of meat) revolved by 498.1 - 758.2 ℃ and the maximum temperature of meat center revolved by 928.5 - 1843.7 ℃, respectively. Whereas the safety margin against DNBR revolved by 6.50 - 2.05. By maximum meat temperature limit of 2804 ℃ and the minimum DNBR of 1.30, it is concluded that the PWR-KSNP design was in the range of safety. (author)

  11. Control rod

    Purpose: To enable semi-permanent and safety use of a control rod in a water cooled type reactor operated under high temperature and high pressure conditions by using a blade in which hafnium materials at a nuclear reactor quality are covered with stainless steels or zirconium alloys. Constitution: A plate-like hafnium material is surrounded with a thin plate of stainless steels or zirconium alloys under vacuum and the joint portions of the thin plate is subjected to seam welding. Then a blade is prepared by welding the remaining joining portions at both ends in a conventional manner. The welding method usable herein includes electron beam welding, laser welding and the like. If it is required to increase the close bondability between the halfnium plate and the thin plate, the blade thus obtained is subjected as it is to extrusion fabrication thereby obtain a desired increased bondability. (Kawakami, Y.)

  12. PWR decontamination feasibility study

    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. PWR type reactor

    From a PWR with a primary circuit, consisting of a reactor pressure vessel, a steam generator and a reactor coolant pump, hot coolant is removed by means of an auxiliary system containing h.p. pumps for feeding water into the primary circuit and being connected with a pipe, originating at the upper part, which has got at least one isolating value. This is done by opening an outlet in a part of the auxiliary system that has got a lower pressure than the reactor vessel. Preferably a water jet pump is used for mixing with the water of the auxiliary system. (orig.)

  14. Investigation, experiment and analysis on PWR sump screen clogging issue

    JNES has been conducting experimental and analytical study to develop an evaluation method concerning the downstream effect of the sump screen clogging issue during LOCA in PWR plants. Flow clogging characteristics were investigated based on data for the relation of pressure loss and flow velocity during flow clogging due to debris accumulation. Deposition of chemical precipitates on the fuel cladding using an electrically heated rod was investigated. A test shows chemical precipitates deposited on the cladding and the deposit was mainly analyzed to be calcium compounds. The analysis with a thermal-hydraulic code on the downstream effect has shown that the core could be cooled because the core inlet flow compensates a evaporation of coolant due to the decay-heat even if core inlet was 99% clogged just after the ECCS recirculation operation started during the cold-leg break LOCA in PWR plants. (author)

  15. Investigation, experiment and analysis on PWR sump screen clogging issue

    JNES has been conducting experimental and analytical study to develop an evaluation method concerning the chemical effect and the downstream effect of the sump screen clogging issue during LOCA in PWR plants. Chemical effect tests show that corrosion of carbon steel and galvanized steal may come to be important in domestic plants, in addition to corrosion of aluminum and insulator which has been considered dominant in the chemical effect. With respect to the downstream effect, deposition of chemical precipitates on the fuel cladding using an electrically heated rod is investigated. A test shows chemical precipitates deposited on the cladding and the deposit was mainly analyzed to be calcium compounds. The analysis on the downstream effect has shown that even if core inlet was completely clogged just after the recirculation operation started during LOCA in PWR plants, although upper part of core may be uncovered temporary and cladding temperature increased, core could be cooled by coolant injection through the hot-leg. (author)

  16. A probabilistic SSYST-3 analysis for a PWR-core during a large break LOCA

    This report demonstrates the SSYST-3 analysis and application for a German PWR of 1300 MW. The report is concerned with the probabilistic analysis of a PWR core during a loss-of-coolant accident due to a large break. With the probabilistic analysis, the distribution functions of the maximum temperatures and cladding elongations occuring in the core can be calculated. Parameters like rod power, the thermohydraulic boundary conditions, stored energy in the fuel rods and the heat transfer coefficient were found to be the most important. The expected value of core damage was determined to be 2.9% on the base of response surfaces for cladding temperature and strain deduced from SSYST-3 single rod results. (orig./HP)

  17. Control rod operation device

    Purpose: To reduce operator's operation burdens in the low power state, while moderate his mental burdens upon high power state for the operation of control rod operation device. Constitution: Coordinate of main control rods to be operated, aimed insertion and withdrawal positions and velocity are calculated by the control rod operation sequence and the control rod worth table, to output a control rod selection signal and a control rod operation signal. The control rod operation device conducts extraction or insertion of control rods by these signals. In this way, the operator can automatically operate the control rods by merely manipulating the control rod operation device, by which the operator's operation burden can be reduced in low power state. Further, since the selection of the control rods, the operation speed, etc. are judged by an electronic computer also upon high reactor power state, operator's metal burdens can be moderated. (Kamimura, M.)

  18. PWR degraded core analysis

    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.)

  19. Recalculation of simulated post-scram core power decay curve for use in ROSA-IV/LSTF experiments on PWR small-break LOCAs and transients

    Simulated post-scram core power decay curve for use in Large Scale Test Facility (LSTF) tests has been calculated on a best-estimate basis, particularly in two points, i.e. estimation of the delayed neutron fission power and consideration of the stored heat in a pressurized water reactor (PWR) fuel rod. The New Power Curve provides a LSTF heater rod with the heat transfer rate from a PWR fuel rod that was estimated for a typical pressure transient during a PWR small-break loss of coolant accident. This approach neglects conservatively the effect of stored heat release from the LSTF heater rod considering that there is large uncertainty in the thermal conductivity of outer insulator in the LSTF heater rod. When the New Power Curve is used as the LSTF core power curve, the heat transfer rate from a LSTF heater rod gives a little conservative values as compared with the heat transfer rate from a PWR fuel rod. (author)

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

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

    1995-08-01

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

  1. A study on thimble plug removal for PWR plants

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

    1997-12-31

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

  2. Axial simulation of PWR core and study of actuators

    Development of an operation code allowing to simulate the behaviour of a PWR type reactor core. Load following is controled by bore and control rods, taking into account the temperature counter-reactions. The fine behaviour of the fuel element during transients is not simulated, on the other hand the central part of the reactor is completely simulated. The regulation equation are easily modifiable and thus it is possible to test in open loop any modification brought about to this regulation. Description of simulation tests on CAS-2B reactor: core control, static tests, dynamic tests

  3. Neutronal aspects of PWR control for transient load following

    The purpose of this thesis is to qualify the CRONOS diffusion code on a load transient in grey mode control. First of all, we have established a general axial calculational model and studied the important physical phenomena: xenon oscillation, grey rods absorption, radial leaks modelling, effect of the initial conditions in Iodine and Xenon. In a second stage, a three dimensional calculation has been performed, the results of which have been compared to a PWR 900 TRICASTIN 3 experiment and have been in good agreement. In the last part, we show that the results of the axial model using one-dimensional CRONOS calculations are quite consistent with the three-dimensional calculation

  4. A flow model for a control rod drop analysis

    In pressurized water reactor (PWR), the core must be shut down quickly to prevent damage of the reactor internals if any operating limits are exceeded. Therefore, the scram time of control rod is one of the most important design parameters for the safety of the nuclear plant. The control rod drop time is affected by various types of fluid resistance force, so the accurate calculation of the flow field will play a significant role in the drop time prediction. In this paper, a new flow model for a control rod drop analysis of PWR is presented. The calculation of flow between rod and thimble tube is described in detail. According to the drop progress of a control rod, the flow analysis is divided into three phases, 1) above the ventilation holes, 2) between the holes and the entrance of dashpot, 3) after the entrance of dashpot. In each phase, several non-linear equations which describe the flow field are established based on the conservation of energy and flow. Then, the Newton iterative method is used to solve these non-linear equations. A computer code is developed based on this model and several sensitivity analyses are carried out by using this code. The effects of structural design parameters changes, namely, the diameter and the length of dashpot, the diameter of the ventilation holes, on the scram time and the fluid resistance force are discussed and presented. These results show that the new model is useful and accurate in the analysis of control rod drop and the code can be an effective tool for the fuel assembly design of PWR. (author)

  5. Heat transfer in rod bundles with severe clad deformations

    The content of the paper is focused on heat transfer conditions during the reflood phase of a LOCA in slightly to severely deformed PWR fuel rod bundle geometries. The status of analytical and, especially, of experimental work is described as far as it is possible within this frame. Emphasis is placed on the presentation of the results of ''Flooding Experiments with Blocked Arrays'' (FEBA), a program performed at the Kernforschungszentrum Karlsruhe in the frame of the Project Nuclear Safety (PNS). (orig./WL)

  6. Analysis of fuel rod behaviour within a rod bundle of a pressurized water reactor under the conditions of a loss of coolant accident (LOCA) using probabilistic methodology

    The assessment of fuel rod behaviour under PWR LOCA conditions aims at the evaluation of the peak cladding temperatures and the (final) maximum circumferential cladding strains. Moreover, the estimation of the amount of possible coolant channel blockages within a rod bundle is of special interest, as large coplanar clad strains of adjacent rods may result in strong local reductions of coolant channel areas. Coolant channel blockages of large radial extent may impair the long-term coolability of the corresponding rods. A model has been developed to describe these accident consequences using probabilistic methodology. This model is applied to study the behaviour of fuel rods under accident conditions following the double-ended pipe rupture between collant pump and pressure vessel in the primary system of a 1300 MW(el)-PWR. Specifically a rod bundle is considered consisting of 236 fuel rods, that is subjected to severe thermal and mechanical loading. The results obtained indicate that plastic clad deformations with circumferential clad strains of more than 30% cannot be excluded for hot rods of the reference bundle. However, coplanar coolant channel blockages of significant extent seem to be probable within that bundle only under certain boundary conditions which are assumed to be pessimistic. (orig./RW)

  7. Reactivity measurements using rod drop and 3D kinetics interpretation in PWRs

    A new method for the PWR control rods worth measurement has been developed and validated within the frame of a joint CEA/EDF/FRAMATOME research program. This method consists in the interpretation of the neutron detectors signals evolution during a rod drop transient. This paper gives the general principles of the method, insisting upon the issues of space and kinetic effects compensation and presents overall results of the on-site experimental validation tests

  8. The PWR programme

    For fueling the PWR type reactors two types of fuel were developed: the UO2 and mixed oxide fuels. To satisfy the demand of the operators of UO2-fuelled power plants a specific industrial organization has been established by Cogema and Framatome: Framagema supplies the technical expertise and sells the fuel; FBFC (Societe Franco-Belge de Fabrication Combustible) is manufacturing the fuel by using particularly the zirconium components produced by Zircotube and Cezus. By making possible the recycling of the materials recovered from the spent fuel reprocessing the MOX (mixed oxide fuels) technology represents an important venture for the future electronuclear sector. To implement this project Cogema created together with Belgonucleaire (the administrator of the Dessel manufacture plant) the GIE COMMOX, in charge with marketing of this fuel. On the other side Cogema which produces MOX in its facility at Cadarache, is at present building the plant at Melox of a capacity of 120 tonnes/year. After presenting the present situation with UO2 and MOX fuels the paper ends with considerations concerning the future fuels and fuels for future and further future reactors

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

    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)

  10. ZZ 3DLWRCT, 3-D LWR Rod Ejection and Rod Withdrawal Benchmarks

    Description: 3-D LWR Rod Ejection and Rod Withdrawal Benchmarks. A - NEA-1398/02 - Description of program or function: - 3DLWRCT-1 represents the first phase of a series of LWR core transient benchmarks. It consists of 2 parts: - PWR problem: ejection of a control assembly from an initially critical core at hot zero power or at full power, each for three different geometrical configurations. The set of problems offers a variety of reactivity excursions which efficiently test the coupled neutronic/thermal-hydraulic models of the codes. - BWR problem: reactivity excursions caused by cold water injection and pressurization events. 2-group macroscopic cross sections and their derivatives respect to boron density, moderator temperature, moderator density and fuel temperature for BWR and PWR core materials are included. For the PWR cases, 63 submitted solutions are analyzed in comparison with a nodal reference solution defined by using a finer spatial and temporal resolution than in standard calculations, which is included. For the BWR benchmark only the cold water injection event is discussed and evaluated in some detail. Lacking a reference solution, the evaluation of the 8 sets of BWR contributions relies on synthetic comparative discussion. - 3DLWRCT-2 represents the second phase of the PWR core transient benchmark and concerns an uncontrolled withdrawal of control rods at zero power. The same cross section data as for phase I is used. B - NEA-1398/03 - Description of program or function: - 3DLWRCT-2 represents the second phase of the benchmark and concerns the results of the 'rod withdrawal from zero power' benchmark, including the solutions obtained by ten participants, from ten different countries, and a reference solution, obtained by refining the spatial and time meshing. The problem is mathematically well defined. The specification provides cross-sections for fuel, reflector and absorbers. Four cases are analysed. The submitted solutions are compared to a

  11. Experimental study for the effects of ballooned rod bundle on the convective heat transfer by single-phase steam flow

    For a large break loss-of-coolant accident (LBLOCA) conditions in a pressurized-water reactor, the cladding temperature increases until the reflood phase and the increased temperature can make a ballooned fuel rods. As a result, the flow passage area of sub-channel is reduced and it leads the redistribution of flow and heat transfer in sub-channels. During the single-phase steam flow in the early phase of the reflood, the cladding temperature may increase and have a peak value due to low heat transfer from the fuel to the steam. If a LBLOCA condition and ballooned fuel rods are occurred, the effect of reduced flow passage on the convective heat transfer by single-phase steam flow is important phenomena to analyze the safety of a reactor. The present experiments were performed in various Reynolds numbers (about 2600∼13000) to investigate the effect of the Ballooned fuel rods on heat transfer phenomena by single-phase steam flow. The experiments were performed in two rod bundles in KAERI reflood ATHER test facility. One is a non-deformed 6x6 rod bundle, which consists of 36 non-deformed heater rods. The other is a deformed 5x5 rod bundle that consists of 9 deformed heater rods and 16 non-deformed heater rods. The cladding temperature and convective heat transfer for two rod bundles are compared for each flow conditions and the effects of experimental parameters are analyzed. (author)

  12. Dynamic modelling of PWR fuel assembly for seismic behaviour

    Vibration and snap back tests have shown that the behaviour of PWR fuel assemblies was non linear : the fuel assembly eigenfrequencies decrease with the excitation level or with the motion amplitude, which was supposed to be due to the slippage of the fuel rods through the grids. Up to now the fuel assembly models were linear and composed by one beam alone representing both the guide thimbles and the fuel rods or by two beams (one for the guide thimbles and one for the fuel rods). The stiffness of such models' were adjusted to fit with the measured eigenfrequency corresponding to a given amplitude. The aim of this paper is to identify the influence of the slippage between grids and fuel rods on the dynamic behaviour of the fuel assembly. For that purpose a non linear fuel assembly model is proposed representing explicitly the slippage phenomenon and is applied to the reduced scale fuel assemblies which have been tested in the framework of a collaboration between FRAMATOME and CEA-DMT. Comparisons between calculations and experiments will be presented and the limitation of this model will be also discussed

  13. 1987 Sucker rod tables

    1987-03-01

    This reference identifies manufacturers qualified to produce API sucker rods and related equipment, lists chemical and mechanical properties of the various types of rods and provides dimensional characteristics. In addition, similar information is given for non-API products such as fiberglass and hollow rods.

  14. Piston rod seal

    Lindskoug, S.

    1984-06-05

    In a piston rod seal of the type comprising a gland through which the piston rod is passed the piston is provided with a sleeve surrounding the piston rod and extending axially so as to axially partly overlap the gland when the piston is in its bottom dead center position. 4 figs.

  15. Piston rod seal

    Lindskoug, Stefan (Malmo, SE)

    1984-01-01

    In a piston rod seal of the type comprising a gland through which the piston rod is passed the piston is provided with a sleeve surrounding the piston rod and extending axially so as to axially partly overlap the gland when the piston is in its bottom dead center position.

  16. Tie rod insertion test

    B. LEVESY

    2002-01-01

    The superconducting coil is inserted in the outer vaccum tank and supported by a set of tie rods. These tie rods are made of titanium alloy. This test reproduce the final insertion of the tie rods inside the outer vacuum tank.

  17. Control rod drives

    Purpose: To prevent damages in control rod drives upon connection and disconnection with control rods by providing, to an extension rod, a closure and opening guide mechanism which is adapted to open and close depending upon connection and disconnection with the control rod. Constitution: In control rod drives having a driving section and a lower mechanism, the lower mechanism has a guide tube engaged into the upper cover of a reactor container and suspended therefrom into the reactor container. An opening and closure mechanism with guide blades capable of contacting the inner wall of the guide tube is secured to an extension rod for fastening the gripper which supports the control rod. Such a mechanism cap prevent damages in the control rod drives and the control rods due to the connection after disconnection of them by the buffering action between the extension rod and the control rod, as well as damages in both of them caused by rolling such as in earthquakes by the buffering action between the extension rod and the guide tube. (Moriyama, K.)

  18. ''SERUS'' an expert system for the ultrasonic examination of fuel rods

    The use of pattern recognition functions and the use of models of the human expert reasoning, allow the automatic identification of defects in welds or structures. The proposed application uses ultrasonic examination to detect and classify 3 types of defects in end plug welds of PWR fuel rods. (author)

  19. Serus, an expert system for the ultrasonic examination of fuel rods

    The use of pattern recognition functions and the modelization of the human expert reasoning, allow the automatic identification of defects in welds or structures. The proposed application uses an ultrasonic examination to detect and classify 3 types of defects in end plug welds of PWR fuel rods

  20. Design of Testing Set-up for Nuclear Fuel Rod by Neutron Radiography at CARR

    WEI; Guo-hai; HAN; Song-bai; WANG; Hong-li; HAO; Li-jie; WU; Mei-mei; HE; Lin-feng; WANG; Yu; LIU; Yun-tao; SUN; Kai; CHEN; Dong-feng

    2012-01-01

    <正>An experimental set-up dedicated to non-destructively test a 15 cm long pressurized water reactor (PWR) nuclear fuel rod by neutron radiography (NR) is designed and fabricated. It consists of three parts: Transport container, imaging block and steel support. The design of the transport container was optimized with Monte-Carlo simulation by the MCNP code.

  1. Sizewell 'B' PWR reference design

    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.)

  2. Analyses of single-phase heat transfer and onset of nucleate boiling in a rod bundle with mixing vane grids

    In the framework of axial offset anomaly risk assessment in Pressurized Water Reactor (PWR) cores, an experimental program involving hydraulic and thermal-hydraulic tests on identical 5x5 bundle geometry was completed. It aimed at developing a consistent set of single-phase heat transfer model and associated onset of nucleate boiling (ONB) wall superheat criterion to further predict the existence and location of boiling zones in a PWR core, using a sub-channel Thermal-Hydraulic (T/H) code. This paper is devoted to the code-based analysis of the experimental data obtained on a bundle equipped with alternating simple support grids and mixing vane grids. Dedicated heat transfer models including a grid enhancement function are developed and the use of Frost & Dzakowic ONB wall superheat criterion is recommended along with these models. (author)

  3. Modelling of pellet-cladding interaction for PWRs reactors fuel rods

    The pellet-cladding interaction that can occur in a PWR fuel rod design is modelled with the computer codes FRAPCON-1 and ANSYS. The fuel performance code FRAPCON-1 analyzes the fuel rod irradiation behavior and generates the initial conditions for the localized fuel rod thermal and mechanical modelling in two and three-dimensional finite elements with ANSYS. In the mechanical modelling, a pellet fragment is placed in the fuel rod gap. Two types of fuel rod cladding materials are considered: Zircaloy and austenitic stainless steel. Linear and non-linear material behaviors are allowed. Elastic, plastic and creep behaviors are considered for the cladding materials. The modelling is applied to Angra-II fuel rod design. The results are analyzed and compared. (author)

  4. Conceptual study of advanced PWR core design

    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

  5. Enriched Gadolinium as burnable absorber for PWR

    This paper is a summary of a master of thesis work in reactor physics made by Ola Seveborn. The work was done at Vattenfall Braensle AB and Ola was guided through the work by the corresponding author of this paper. The results presented are calculations for Ringhals 3, which is a Westinghouse 3-loop PWR within the Vattenfall Group. The fuel is characterized by 17x17 assemblies of AFA type containing 3.80-3.95 w/o 235U and 8 rods containing 2 w/o Gadolinium with an enrichment of 70 w/o 157Gd. The calculations were performed with the Studsvik-Scandpower code package based on the CASMO-4 lattice code and the SIMULATE-3 nodal code. The results are compared to the corresponding calculations for fuel with 5 w/o gadolinium with natural isotopic constitution. The depletion of the cores was done separately for the reference and enriched case. The results show that the gains in average for the five cycles studied are about 70 EFPH per cycle. This is an effect of the lower gadolinium content needed. Also less parasitic absorption of enriched gadolinium in the end of the fuel life contributes to the increased cycle lengths. The abruptly increased reactivity and internal power peaking factor around 10 MWd/kgU do not affect the core design negatively. (authors)

  6. Conceptual study on advanced PWR system

    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

  7. Conceptual study of advanced PWR core design

    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.

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

    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.)

  9. Sucker rod construction

    Anderson, R.A.; Goodman, J.L.; Tickle, J.D.; Liskey, A.K.

    1987-03-31

    A sucker rod construction is described comprising: a connector member being formed to define a rod receptacle having a closed axially inner end and an open axially outer end, the rod receptacle having axially spaced, tapered annular surfaces, a cylindrical fiberglass rod having an end having an outer surface being received within the rod receptacle through the outer end and cooperating therewith to define an annular chamber between the outer surface of the end of the rod and the tapered annular surfaces, and a bonding means positioned in the annular chamber for bonding to the outer surface of the end of the rod to confront the tapered annular surfaces, each annular surface having an angle of taper with respect to the outer surface of the fiberglass rod, and each angle of taper being progressively and uniformly less toward the open end by an amount between one and one-half degrees and two degrees, inclusive, and a collet connected to the connector member adjacent the open axially outer end of the rod receptacle and having an axial bore therethrough retaining the end of the rod in coaxial position within the rod receptacle.

  10. Sucker rod guide

    Edwards, B.J.; Starks, J.A.

    1989-08-22

    This patent describes a sucker rod guide for mounting on a sucker rod and spacing the sucker rod from the tubing in an oil well. The guide comprising a generally cylindrically-shaped, extruded, ultra-high density polyethylene body having a substantially smooth outside surface; a longitudinal bore provided centrally of the body. The bore having a smaller diameter than the diameter of the sucker rod; a plurality of grooves provided in circumferential relationship in the bore; and a tapered slot extending longitudinally through the body from the outside surface to the bore. The tapered slot further comprising a slot mouth located at the outside surface and a slot throat spaced from the slot mouth. The slot throat lying adjacent to the sucker rod bore and wherein the slot throat is wider than the slot mouth for mounting the sucker rod guide on the sucker rod.

  11. A Novel Burnable Absorber Concept for PWR: BigT (Burnable Absorber-Integrated Guide Thimble)

    This paper presents the essential BigT design concepts and its lattice neutronic characteristics. Neutronic performance of a newly-proposed BA concept for PWR named BigT is investigated in this study. Preliminary lattice analyses of the BigT absorber-loaded WH 17x17 fuel assembly show a high potential of the concept as it performs relatively well in comparison with commercial burnable absorber technologies, especially in managing reactivity depletion and peaking factor. A sufficiently high control rod worth can still be obtained with the BigT absorbers in place. It is expected that with such performance and design flexibilities, any loading pattern and core management objective, including a soluble boron-free PWR, can potentially be fulfilled with the BigT absorbers. Future study involving full 3D reactor core simulations with the BigT absorbers shall hopefully verify this hypothesis. A new burnable absorber design for Pressurized Water Reactor (PWR) named 'Burnable absorber-Integrated control rod Guide Thimble' (BigT) was recently proposed. Unlike conventional burnable absorber (BA) technologies, the BigT integrates BA materials directly into the guide thimble but still allows insertion of control rod (CR). In addition, the BigT offers a variety of design flexibilities such that any loading pattern and core management objective can potentially be fulfilled

  12. A Novel Burnable Absorber Concept for PWR: BigT (Burnable Absorber-Integrated Guide Thimble)

    Yahya, Mohdsyukri; Kim, Yonghee [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Chung, Chang Kyu [KEPCO Engineering and Construction Company, Daejeon (Korea, Republic of)

    2014-05-15

    This paper presents the essential BigT design concepts and its lattice neutronic characteristics. Neutronic performance of a newly-proposed BA concept for PWR named BigT is investigated in this study. Preliminary lattice analyses of the BigT absorber-loaded WH 17x17 fuel assembly show a high potential of the concept as it performs relatively well in comparison with commercial burnable absorber technologies, especially in managing reactivity depletion and peaking factor. A sufficiently high control rod worth can still be obtained with the BigT absorbers in place. It is expected that with such performance and design flexibilities, any loading pattern and core management objective, including a soluble boron-free PWR, can potentially be fulfilled with the BigT absorbers. Future study involving full 3D reactor core simulations with the BigT absorbers shall hopefully verify this hypothesis. A new burnable absorber design for Pressurized Water Reactor (PWR) named 'Burnable absorber-Integrated control rod Guide Thimble' (BigT) was recently proposed. Unlike conventional burnable absorber (BA) technologies, the BigT integrates BA materials directly into the guide thimble but still allows insertion of control rod (CR). In addition, the BigT offers a variety of design flexibilities such that any loading pattern and core management objective can potentially be fulfilled.

  13. Condensate purification in PWR reactors

    The recommendations made by the VGB task group on 'condensate purification for PWR reactors' 1976 are discussed in detail. Techniques and circuiting possibilities of condensate purification for BBR steam generators (forced circulation) and KWU steam generators (U tube with blow-down) are mentioned. (HP)

  14. PWR AXIAL BURNUP PROFILE ANALYSIS

    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)

  15. PWR AXIAL BURNUP PROFILE ANALYSIS

    J.M. Acaglione

    2003-09-17

    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&W 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).

  16. Dynamic rod worth measurements (''Rod Insertion''). Final report for the period 01 December 1994 - 30 November 1996

    Reload startup physics tests are performed for pressurized water reactors (PWR power plant) following a refuelling or other significant core alteration for which nuclear design calculations are required. Part of the reload startup physics tests are control rod group worths measurements. for this purpose a new so-called method ''Rod-Insertion'' was developed. It can also be used as an additional measuring instrument on the research reactor for education purposes. The principle of the rod-insertion method is to start from a critical reactor operating at low power and to measure the time-dependent reactivity change while a control rod is inserted into the core. Unlike in the rod-drop method, the measured control rod is inserted with the drive mechanism at normal speed. By analyzing the flux trace using point-kinetics, not only the total rod worth but also the differential and the integral rod worth curves are obtained. A high-quality electrometer is required for monitoring the neutron flux. The analysis is performed by transferring the data to an IBM PC compatible with some additional standard electronic board and the associated software. The new reactivity meter has been validated on the TRIGA Mark II reactors in Ljubljana and Vienna and at the Krsko Nuclear Power Plant during physics startup tests after reload. The results proved the high performance of the reactivity meter in the standard applications according to the existing procedures, as well as in the new rod-insertion technique of measuring the control rod group worths. This method drastically differs from others such as absence of any chemical control of reactivity (like boron exchange method), and minimizing a testing time and waste coolant production

  17. Investigation of the conservatism in the traditional approach to calculating the number of fuel rods in DNB

    The purpose of this paper is to describe a statistical method of analysis used to generate a distribution function for the number of fuel rods in DNB. The method is modified and applied to a typical 4 loop PWR plant operating at a limiting DNB condition, and at locked rotor and single rod withdrawal fault conditions. It is concluded that the traditional approach to calculating the number of fuel rods in DNB, which counts all fuel rods with DNBRs less than the limit of the DNB correlation as failed, is conservative for safety analyses

  18. Experimental analysis of redistribution of the transversal crossflow in rod bundles

    Fuel elements for PWR type nuclear reactors consist of rod bundles, in a square array and are held by grids. The coolant flows, mainly, axially along the rods. The inlet flow bad distribution can yield a strong crossflow. The present work consists in the experimental analysis of the transversal crossflow between 2 bundles with 4x4 rods each, with and without the presence of spacer-type grids, for several inlet flow conditions. It was observed that the crossflow is strongly dependent of the static pressure difference between the bundles and that the presence of grids induces a rapid homogenization of the flow. (C.M.)

  19. An analytical model for the prediction of fluid-elastic forces in a rod bundle subjected to axial flow: theory, experimental validation and application to PWR fuel assemblies; Calcul des forces fluidelastiques dans les faisceaux de tubes sous ecoulement axial: theorie, validation, application aux assemblages combustibles des REP

    Beaud, F. [Electricite de France (EDF), 78 - Chatou (France)

    1997-12-31

    A model predicting the fluid-elastic forces in a bundle of circular cylinders subjected to axial flow is presented in this paper. Whereas previously published models were limited to circular flow channel, the present one allows to take a rectangular flow external boundary into account. For that purpose, an original approach is derived from the standard method of images. This model will eventually be used to predict the fluid-structure coupling between the flow of primary coolant and a fuel assemblies in PWR nuclear reactors. It is indeed of major importance since the flow is shown to induce quite high damping and could therefore mitigate the incidence of an external load like a seismic excitation on the dynamics of the assemblies. The proposed model is validated on two cases from the literature but still needs further comparisons with the experiments being currently carried out on the EDF set-up. The flow has been shown to induce an approximate 12% damping on a PWR fuel assembly, at nominal reactor conditions. The possible grid effect on the fluid-structure coupling has been neglected so far but will soon be investigated at EDF. (author). 16 refs.

  20. Fuel rod bowing

    The purpose of this investigation was to quantify the extent of fuel rod bowing in Westinghouse pressurized water reactors and to assess the effects of fuel rod bowing on plant safety and reliability. An empirical bow correlation was developed based on data from irradiated assemblies. Analyses conducted with these conservative empirical predictions show that: (1) generically identified DNBR margins are adequate to offset DNBR reductions due to rod bow, (2) the present design practice of increasing the highest calculated core peaking factor is sufficient to account for all deviations, including the effects of rod bow, and (3) fretting and corrosion of bowed rods are negligible. These conclusions indicate that fuel rod bowing results in no impact on plant safety or reliability

  1. Reactor control rod

    Object: To enable quick descent of a control rod body even when some relative phase deviation between upper drive means and wrapper tube is produced, while permitting a coolant to effectively flow into a protective tube irrespective of the position of the control rod body. Structure: In a control rod used for a nuclear reactor such as a fast breeder, an orifice which dispenses with a cylindrical guide tube and has a greater inner diameter than the outer diameter of the protective tube of the control rod body is provided on the inner side of a wrapper tube, thus permitting smooth operation of the control rod body and also permitting the coolant to effectively flow into the protective tube irrespective of the control rod body. (Horiuchi, T.)

  2. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme

  3. Status of rod consolidation

    Two of the factors that need to be taken into account with rod consolidation are (1) the effects on rods from their removal from the fuel assembly and (2) the effects on rods as a result of the consolidation process. Potential components of both factors are described in the report. Discussed under (1) are scratches on the fuel rod surfaces, rod breakage, crud, extended burnup, and possible cladding embrittlement due to hydrogen injection at BWRs. Discussed under (2) are the increased water temperature (less than 100C) because of closer packing of the rods, formation of crevices between rods in the close-packed mode, contact with dissimilar metals, and the potential for rapid heating of fuel rods following the loss of water from a spent fuel storage pool. Another factor that plays an important role in rod consolidation is the cost of disposal of the nonfuel-bearing components of the fuel assembly. Also, the dose rate from the components - especially Inconel spacer grids - can affect the handling procedures. Several licensing issues that exist are described. A list of recommendations is provided. 98 refs., 5 figs., 5 tabs

  4. Dynamic Rod Worth Measurement

    The dynamic rod worth measurement (DRWM) technique is a method of quickly validating the predicted bank worth of control rods and shutdown rods. The DRWM analytic method is based on three-dimensional, space-time kinetic simulations of the rapid rod movements. Its measurement data is processed with an advanced digital reactivity computer. DRWM has been used as the method of bank worth validation at numerous plant startups with excellent results. The process and methodology of DRWM are described, and the measurement results of using DRWM are presented

  5. Drilling Experiments of Dummy Fuel Rods Using a Mock-up Drilling Device and Detail Design of Device for Drilling of Irradiated Nuclear Fuel Rods

    Kim, Jae Yong; Lee, H. K.; Chun, Y. B.; Park, S. J.; Kim, B. G

    2007-07-15

    KAERI are developing the safety evaluation method and the analysis technology for high burn-up nuclear fuel rod that is the project, re-irradiation for re-instrumented fuel rod. That project includes insertion of a thermocouple in the center hole of PWR nuclear fuel rod with standard burn-up, 3,500{approx}4,000MWD/tU and then inspection of the nuclear fuel rod's heat performance during re-irradiation. To re-fabricate fuel rod, two devices are needed such as a drilling machine and a welding machine. The drilling machine performs grinding a center hole, 2.5 mm in diameter and 50 mm in depth, for inserting a thermocouple. And the welding machine is used to fasten a end plug on a fuel rod. Because these two equipment handle irradiated fuel rods, they are operated in hot cell blocked radioactive rays. Before inserting any device into hot cell, many tests with that machine have to be conducted. This report shows preliminary experiments for drilling a center hole on dummy of fuel rods and optimized drilling parameters to lessen operation time and damage of diamond dills. And the design method of a drilling machine for irradiated nuclear fuel rods and detail design drawings are attached.

  6. Modeling of fuel-rod behavior during reactor power cycling and ramping experiments with computer code FRAPCON-2

    Modelling of fuel-rod behavior during reactor power cycling and ramping (including power-cooling mismatch experiments) with the computer code FRAPCON-2 is discussed. FRAPCON-2 computer calculations, using different mechanical models (Rigid Pellet, Deformable Pellet and Finite Element Mechanical Models) are compared with experimental results. The range of conditions over which FRAPCON-2 may be applied for PWR fuel rod behavior modelling during reactor power cycling and ramping are illustrated

  7. Overview of PWR chemistry options

    Nordmann, F.; Stutzmann, A.; Bretelle, J.L. [Electricite de France, Central Labs. (France)

    2002-07-01

    EDF Central Laboratories, in charge of engineering in chemistry, of defining the chemistry specifications and studying the operation feedback and improvement for 58 PWR units, have the opportunity to evaluate many options of operation developed and applied all around the world. Thanks to these international relationships and to the benefit of a large feedback from many units, some general evaluation of the various options is discussed in this paper. (authors)

  8. Optimal burnable poison-loading in a PWR with carbon coated particle fuel

    An innovative PWR concept that uses carbon-coated particle fuels moderated by graphite as that of HTGR but cooled by pressurized light water has been studied. The aim of this concept is to take both the best advantages of fuel integrity against fission products release and the reliability PWR technology based on the long operational experience. The purpose of the study is to optimize loading pattern of burnable poison in the proposed core in order to suppress excess reactivity during a cycle. Although there are many parameters to be determined for optimization of the usage of burnable poison, the emphasis is put here on loading patterns of Gadolinia in an assembly and in the core. We investigated the burnup characteristics of the core varying the concentration of burnable poison in a fuel rod, the number of burnable poison-rods in an assembly, and the number of burnable poison-assemblies in the core. The result suggested that Gadolinia was more suitable for this reactor than boron as burnable poison, and it was possible to make the reactivity swing negligible by combining at least three kinds of burnable poison-assemblies in which the amount of Gadolinia was different. Therefore the requirement for the number of control rods was reduced and it meant that Control Rod Programming would become easier. (author)

  9. Assessment of the interchannel mixing model with a subchannel analysis code for BWR and PWR conditions

    The influence of the interchannel mixing model employed in a traditional subchannel analysis code was investigated in this study, specifically on the analysis of the enthalpy distribution and critical heat flux (CHF) in rod bundles in BWR and PWR conditions. The equal-volume-exchange turbulent mixing and void drift model (EVVD) was embodied to the COBRA-IV-I code. An optimized model of the void drift coefficient has been devised in this study as the result of the assessment with the two-phase flow distribution data for the general electric (GE) 9-rod and Ispra 16-rod test bundles. The influence of the subchannel analysis model on the analysis of CHF was examined by evaluating the CHF test data in rod bundles representing PWR and BWR conditions. The CHFR margins of typical light water nuclear reactor (LWR) cores were evaluated by considering the influence on the local parameter CHF correlation and the hot channel analysis result. It appeared that the interchannel mixing model has an important effect upon the analysis of CHFR margin for BWR conditions. (orig.)

  10. Modeling of the thermo-mechanical behaviour of the PWR fuel

    This article reviews the various physical phenomena that take place in an irradiated fuel rod and presents the development of the thermo-mechanical codes able to simulate them. Though technically simple the fuel rod is the place where appear 4 types of process: thermal, gas behaviour, mechanical and corrosion that combine involving 5 elements: the fuel pellet, the fuel clad, the fuel-clad gap, the inside volume and the coolant. For instance the pellet is the place where the following mechanical processes took place: thermal dilatation, elastic deformation, creep deformation, densification, solid swelling, gaseous swelling and cracking. The first industrial code simulating the behaviour of the fuel rod was COCCINEL, it was developed by AREVA teams from the American PAD code that was included in the Westinghouse license. Today the GALILEO code has replaced the COPERNIC code that was developed in the beginning of the 2000 years. GALILEO is a synthesis of the state of the art of the different models used in the codes validated for PWR and BWR. GALILEO has been validated on more than 1500 fuel rods concerning PWR, BWR and specific reactors like Siloe, Osiris, HFR, Halden, Studsvik, BR2/3,...) and also for extended burn-ups. (A.C.)

  11. Assessment of Inner Channel Blockage on the Annular Fuel Rod

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

    2010-10-15

    A dual-cooled annular fuel for a pressurized water reactor (PWR) has been introduced for a significant amount of reactor power uprate. The Korea Atomic Energy Research Institute (KAERI) has been performing a research to develop a dual-cooled annular fuel for the power uprate of 20% in an optimized PWR in Korea, OPR1000. An inner channel blockage is principal one of technical issues of the annular fuel rod. The inner channel in an annular fuel is isolated from the neighbor channels unlike the outer channels. The inner channel will be faced with a DNB accident by the partial blockage. In this paper, the largest fractional channel blockage was assessed by subchannel analysis code MATRA-AF and an end plug design to complement inlet blockage of inner channel was estimated by CFD code, CFD-ACE

  12. PWR secondary water chemistry guidelines: Revision 3

    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)

  13. Rod drop measurement analysis

    In some cases control rod worth efficiencies evaluated by inverse point kinetics from out-core detector currents remarkable differ from direct calculations. Explanation of this effects is given and is supported by the analysis of some WWER-440 rod drop experiments. (Authors)

  14. Siemens Nuclear Power Corporation experience with BWR and PWR fuels

    The large data base of fuel performance parameters available to Siemens Nuclear Power Corporation (SNP), and the excellent track record of innovation and fuel reliability accumulated over the last twenty-three years, allows SNP to have a clear insight on the characteristics of future developments in the area of fuel design. Following is a description of some of SNP's recent design innovations to prevent failures and to extend burnup capabilities. A goal paramount to the design and manufacture of BWR and PWR fuel is that of zero defects from any case during its operation in the reactor. Progress has already been made in achieving this goal. This paper summarized the cumulative failure rate of SNP fuel rod through January 1992

  15. Shutdown Chemistry Process Development for PWR Primary System

    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.

  16. Conceptual design of SPWR, a PWR with enhanced passive safety

    A conceptual design has been carried out on a new type of integrated pressurized water reactor, SPWR (System-integrated PWR). This reactor installs a poison tank (borated water filled) in the reactor vessel instead of control rod drive system. Three hydraulic pressure valves are installed as the upper interface between the poison tank and primary coolant. A 700MWe power plant with twin 1100MWt SPWRs which are installed in a reactor building has been studied. Design and analysis have been made on the reactor core, reactor (reactor vessel, steam generator, main circulating pump, pressurizer, poison tank and their integration), plant systems (main and sub systems), layout, construction scheme, operation and maintenance, safety related components, reactor dynamics, economics and R and D needs. Passive safety features are also studied. (author)

  17. PWR-blowdown heat transfer separate effects program

    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.

  18. Influence of spectral history on PWR full core calculation results

    The few-group cross section libraries, used by reactor dynamics codes, are affected by the spectral history effect-a dependence of fuel cross sections not only on burnup, but also on local spectral conditions during burnup. A cross section correction method based on Pu-239 concentration was implemented in the reactor dynamic code DYN3D. This paper describes the influence of a cross section correction on full-core calculation results. Steady-state and burnup characteristics of a PWR equilibrium cycle, calculated by DYN3D with and without cross section corrections, are compared. A study has shown a significant influence of spectral history on axial power and burnup distributions as well as on calculated cycle length. An impact of the correction on transient calculations is studied for a control rod ejection example. (Authors)

  19. Rod Photoreceptors Detect Rapid Flicker

    Conner, J. D.; MacLeod, Donald I. A.

    1977-01-01

    Rod-isolation techniques show that light-adapted human rods detect flicker frequencies as high as 28 hertz, and that the function relating rod critical flicker frequency to stimulus intensity contains two distinct branches. (MLH)

  20. Application of the ballooning analysis code MATARE on a generic PWR fuel assembly

    The MATARE (MAbel-TAlink-RElap) code is a new multi-pin deformation analysis code created through the dynamic coupling between the thermal-hydraulic code RELAP5 and multiple instances of the single-pin thermal-mechanics code MABEL. A multi-pin representation of different zones of a typical PWR fuel assembly under post-LOCA reflooding conditions was analysed including some of the most relevant features that characterise a typical nuclear reactor fuel assembly and evaluate their effect on the behaviour of the fuel rods under conditions leading to clad ballooning. The code was able to simulate the deformation of wide regions of a fuel assembly under reflood conditions and has shown how differences in pin pressure and the presence of rod with burnable poisons and control rod guide thimbles also contribute to a substantial incoherent ballooning in agreement with the experimental data. (author)

  1. Modifications needed to operate PWR's plants in G-Mode

    The production of electricity from PWR nuclear plants represents 44% of the total production of electricity in France for 1984, and 68% of the electricity produced by Thermal power plants (127 TWh over 187 TWh). These data show clearly that the French PWR plants do not work in ''base mode'' anymore but have to fit production with consumption, in other words to assume the frequency control. To participate permanently to the load follow and frequency control, it appeared that some improvements in the field of pressurizer level and pressure control were necessary as well as in the field of operator aids computer. It should be noted that these improvements are useful even without taking into account the constraints due to load follow and frequency control because of the mechanical stress in the CVCS piping, for instance. Some additional tests are planned to better identify this specific problem. The need of a more flexible operating mode than ones given by the initial system (black control rods), significantly reduced in 1973 due to the application of the ECCS criterion, led EDF and Framatome to develop a new operating mode (G. Mode) allowing a faster power escalation (5% PN/mn) whatever the fuel burn-up. This new operating mode improves significantly also the flexibility of operation when the frequency control is needed, and helps a lot the operators in such cases. All the 900 MWe Nuclear plants will be able to operate in ''G mode'' before the end of 1984

  2. DNB analysis with mechanistic models for PWR fuel assemblies

    In order to predict the DNB heat flux of PWR fuel assemblies and the critical power of BWR fuel bundles, the Boiling Transition Analysis Code CAPE' has been developed in the IMPACT project. The CAPE code for PWR includes three analysis modules, subchannel analysis module, three-dimensional two-phase flow analysis module, and DNB evaluation module. The subchannel module uses drift-flux model to identify the hottest subchannel. The three-dimensional two-phase flow analysis module uses nonhomogeneous and nonequilibrium two fluid model to analyze the detailed three-dimensional two-phase flow behaviors such as void distribution. For DNB heat flux prediction, the DNB evaluation module uses the Weisman model in which is a mechanistic DNB evaluation model. This paper describes the analysis models, analysis techniques and the results of validation by rod bundle test analysis. To date, the average difference between calculated and 11 measured values was -0.6% with a standard deviation of 7.0%. (author)

  3. Control rod assembly

    In a control rod assembly comprising an extension rod extended upwardly from the upper end of a control rod main body disposed in a reactor core and an extension tube engaging a grip portion disposed to an upper portion of the extension rod for suspending the control rod main body, a shrinkable portion is disposed to a part of the extension tube or extension rod, or a grip portion shrinkable in the axial direction is disposed to the extension rod. Further, a spring is interposed to a portion of the extension tube and bellows are disposed to the inner side or the outer side of the spring. A double-cylindrical temperature sensing member is disposed surrounding the outer side of the bellows or the spring. Liquid metals are sealed in the temperature sensing member or the bellows. This can improve the response of the coolants to the temperature elevation and can suppress the change of the reactor core insertion amount relative to temperature change during usual operation. (T.M.)

  4. Experimental study on critical heat flux with long rod bundle

    Some new structural features of a long rod bundle test section with experimental method of a type appropriate to the circumstances, data processing and research results are described. The experiments have been performed at a high pressure heat transfer loop. The arrangement of the rod bundle is 3 x 3, square. The rod diameter is 9.5mm, pitch 12.6mm. The effective heating length of each rod is 2200mm. Water or steam-water mixture flows through the rod bundle upwards. The axial heat flux profiles are uniformly distributed. Parameter ranges for the experiments are pressures from 14.4 to 15.7MPa, mass velocities from 1200 to 3540kg/(m2·s) and burnout qualities from -17% to +15%. The data processing for a total of 95 CHF data points is carried out on VAX computer. The formula for predicting CHF values presened can be used in a field of thermal hydraulic design and safety analysis for PWR

  5. Ballooning analysis for the Sizewell B PWR using symmetric MABEL calculations

    An analysis of the fuel clad ballooning potential associated with the Sizewell B PWR following a design basis large break cold leg LOCA is described. Calculations employ MABEL-2C code. No allowance has been made for asymmetries in power or geometry, thus precluding any amelioration offered by early clad rupture. Thermal hydraulic data were derived from a TRAC-PD2 best estimate analysis of the LOCA and the work includes a detailed sensitivity study which leads to a correlation between peak clad temperature and clad strain. For the best estimate start of cycle 1 peak rod rating, no loss of coolability is expected within 95 percent confidence limits on peak clad temperature. No loss of coolability is expected either for rods at the design basis peak rod rating. The temperature does not have to be much higher than the 95 percent confidence limit on the best estimate rating or much beyond that of the design basis rating for rod contact and severe blockage to follow. This indicates that to establish a complete safety case the added complexity of pellet eccentricity and rod to rod power variations must be considered. (U.K.)

  6. A modified statistical methodology for rod internal pressure calculation

    An existing statistical methodology for the nuclear fuel rod internal pressure(RIP) of the Korean PWR fuel has been modified in order to both reduce over-conservatism of the current KAERI deterministic methodology and simplify the complicated procedure of the existing statistical methodology which employs the response surface method and Monte Carlo simulation. The modified statistical methodology employs the system moment method combined with a deterministic approach in determining a maximum variance of RIP distribution. This approach makes the modified statistical methodology much more efficient in the routine reload design analysis since it eliminates the numerous processes required for the power history-dependent RIP variance derivations

  7. Utility implementation of EPRI rod ejection accident methodology

    This report describes the application of ARROTTA, a three dimensional space time kinetics code, to a licensing analysis of the PWR rod ejection accident. Three approaches for the use of ARROTTA are described: (1) a benchmark for point kinetics, (2) direct application as a biased licensing model, and (3) as a best estimate model used in conjunction with statistical combination of uncertainties. The use of ARROTTA as a biased licensing model was fully developed in conjunction with Duke Power Company; the results have been submitted to NRC as part of their reload licensing methodology

  8. Behavior of irradiated PWR fuel under a simulated RIA condition. Results of NSRR Test MH-3

    Results from the power burst experiment, Test MH-3, conducted in the Nuclear Safety Research Reactor (NSRR) are presented. A fuel rod was irradiated with a fuel burnup up to 38.9 MWd/kgU in the Mihama unit No.2 of the Kansai Electric Power Co., Inc. The Test MH-3 was the third and final experiment in a reactivity initiated accident (RIA) test series with the MH fuel rod. Data concerning test method, pre-pulse examination, pulse irradiation, transient records and post-pulse fuel examination are described, and discussions of the results are presented. A test fuel rod is a short-sized 14x14 pressurized water reactor (PWR) type rod, which is refabricated from a full size commercial fuel rod. A double container-type capsule contains the instrumented test fuel rod with stagnant water cooling condition at atmospheric pressure and ambient temperature. The test fuel rod was subjected to the pulse irradiation resulting in a deposited energy of 87 cal/g·fuel and a peak fuel enthalpy of 67 cal/g·fuel. Behavior of the test fuel rod was assessed from pre- and post-pulse examinations and transient records during the pulse irradiation. Cladding surface temperature increased about 200degC. The maximum cladding deformation was 1.6% and the test fuel rod did not fail. Estimated fission gas release during the pulse irradiation is 3.8% in Kr, and 2.3% in Xe, respectively. Through the detail fuel examination, information concerning microstructural change in the fuel pellets was also obtained. (author)

  9. Neutronic Analysis of Advanced SFR Burner Cores using Deep-Burn PWR Spent Fuel TRU Feed

    In this work, an advanced sodium-cooled fast TRU (Transuranics) burner core using deep-burn TRU feed composition discharged from small LWR cores was neutronically analyzed to show the effects of deeply burned TRU feed composition on the performances of sodium-cooled fast burner core. We consider a nuclear park that is comprised of the commercial PWRs, small PWRs of 100MWe for TRU deep burning using FCM (Fully Ceramic Micro-encapsulated) fuels and advanced sodium-cooled fast burners for their synergistic combination for effective TRU burning. In the small PWR core having long cycle length of 4.0 EFPYs, deep burning of TRU up to 35% is achieved with FCM fuel pins whose TRISO particle fuels contain TRUs in their central kernel. In this paper, we analyzed the performances of the advanced SFR burner cores using TRU feeds discharged from the small long cycle PWR deep-burn cores. Also, we analyzed the effect of cooling time for the TRU feeds on the SFR burner core. The results showed that the TRU feed composition from FCM fuel pins of the small long cycle PWR core can be effectively used into the advanced SFR burner core by significantly reducing the burnup reactivity swing which reduces smaller number of control rod assemblies to satisfy all the conditions for the self controllability than the TRU feed composition discharged from the typical PWR cores

  10. Evaluation of the presence of a burnable absorber in an assembly 3x3 type PWR; Evaluacion de la presencia de un absorbedor quemable en un ensamble 3x3 tipo PWR

    Martinez F, M. A.; Del Valle G, E.; Alonso V, G. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, Mexico D. F. 07738 (Mexico)]. e-mail: mike_ipn_esfm@hotmail.com

    2008-07-01

    In the present work the effect is evaluated that causes the presence of a burnable absorber in an adjustment of rods of 3x3 of a fuel assembly type PWR using CASMO-4 code, when comparing the infinite multiplication factor and some average cross sections by means of codes MCNP-4A, CASMO-3 and HELIOS. For this evaluation two cases are evaluated: first consists of an adjustment of rods of 3x3 full completely of fuel and the second consists of a central rod full with a burnable absorber type wet annular burnable absorber (WABA) and the remaining full fuel rods. In both cases the enrichment of the fissile isotopes is varied, for two types of fuel, MOX degree armament and UO{sub 2}. (Author)

  11. Thermodynamic modelling of PWR coolant

    Spinel solubilities on PWR primary circuit surfaces vary with temperature, pH and coolant H2 concentration. The available solubility data are discussed for Fe, Ni, Co and Zn oxides, and species are identified where data are very limited or absent. An equilibrium thermodynamic model is described to predict the solubility, and results are described predicting relative Fe and Ni solubility under normal operating conditions and during shutdown/startup. The relative stabilities of stoichiometric and non-stoichiometric zinc ferrite spinels are also considered. (R.P.)

  12. Flexible sucker rod unit

    Allen, L.F.

    1987-02-03

    This patent describes a deep well having: a. an education tube with an inside diameter extending from the surface of the earth to far below the surface, b. a reciprocating pump housing attached to the bottom of the education tube, c. pump jack means at the surface for reciprocating the pump, d. a light sucker rod connected to the pump jack means and extending into the education tube, and e. a series of heavy sinker bars having a large cross sectional area in the education tube connecting the light sucker rod to the pump; f. an improved integral metal flexible rod unit interconnecting the sinker bars comprising in combination with the above: g. a coupling on each end of the integral metal flexible rod unit connecting the flexible rod unit to the contiguous sinker bar, h. a segment which is flexible as compared to the sinker bars connecting one of the couplings to i. an integral metal bearing adjacent to the other of the couplings, the bearing having j. a cylindrical surface with k. a diameter i. only slightly smaller than the inside diameter of the education tube thereby forming a sliding fit therewith, and ii. greater than the diameter of any other portion of the flexible rod unit and the sinker bar, and l. grooves in the cylindrical surface for the passage of fluid between in the education tube around the bearing.

  13. Control rod drives

    Purpose: To rapidly detect the position to which a control rod has been rapidly inserted into the reactor core upon scram in the control rod drives for use in LMFBR type reactors. Constitution: In control rod drives comprising an acceleration spring disposed to the outside of an extension pipe and an acceleration pipe for conducting the spring force to a control rod for rapidly dropping the rod into the reactor core, a magnet having a repulsive force is disposed to each acceleration pipe and guide pipe as decelerating and buffering means for the acceleration pipe. The position of the control rod is detected by the interaction between the magnet and the coils attached to the inside of the guide pipe or reactor lead switch. According to this invention, 85 % scram signal which has hitherto been difficult to be processed electrically can be obtained with a sufficient intensity and with no delay to thereby improve the entire safety of the reactor system. Then, the inserted position and the insertion time can accurately and rapidly be detected. (Horiuchi, T.)

  14. Control rod drives

    Purpose: To fix a magnetic rotor to a drive shaft and at the time of non-driving, to restrain the rotor by permanent magnets thereby to hold the position of the control rod safely and accurately. Constitution: A control rod position holding device is provided in a motor or a drive shaft of a control rod drive. This device consists of a rotor and a stator, the rotor being provided on its circumference with salient-poles arranged equidistantly, and the position of the rotor being determined depending upon the transfer distance of the control rod and the conversion ratio of the converter. On the other hand, the stator has salient-poles (any of them is a permanent magnet) having the number of poles and the positional relationship equivalent to those of the rotor, and provided in the inner periphery of a cylinder using the drive shaft as a central shaft and wound with a winding. When the control rod is not driven, the poles of the rotor are attracted by the magnetic force of the confronting poles of the stator, thereby to prevent the inverse rotation of the motor shaft due to the dead weight of the control rod. When a current is caused to flow through the winding, the magnetic force of the permanent magnet, and the stator release the rotor. (Yoshino, Y.)

  15. Control rod blocking device

    Purpose: To increase the degree of freedom for the reactor operation by control rod blocking by monitoring the critical power ratio (CPR) with real time. Constitution: There has been a problem that the withdrawal of control rods may occasionally be inhibited with all the margin in view of CPR. The present invention dissolves this problem. That is, the control rod withdrawal device periodically calculates CPR, and calculated CPR upon generation of a control rod withdrawing signal by conpensating the result of calculation with a LPRM signal and a reactor core flow rate signal. The CPR at real time is compared with a predetermined setting value to output a control rod withdrawing inhibition signal depending on the result of the comparison. In the device as described above, since CPR is monitored at real time, the control rod can be withdrawn without causing fuel damages, as well as the inhibition of withdrawal irrespective of the presence of margin in view of CPR can be avoided. Accordingly, degree of freedom in the reactor operation can be increased. (Kamimura, M.)

  16. Evaluation of phenomenological DNB models for rod bundle geometries

    Seven phenomenological DNB models based on the liquid sublayer dryout and the bubble crowding mechanisms have been evaluated for the square array rod bundles at PWR conditions. The local thermal hydraulic conditions were calculated by the COBRA-IV-I code, and it was assumed that the enthalpy and mass velocity distribution in the test bundle would not be changed at different power levels. A simplified method, which has been proposed for the prediction of CHF in rod bundles from round tube CHF correlations without detailed subchannel analyses, was also applied with the phenomenological models. In view of the prediction accuracy and the applicable range, Lin model shows the best result among phenomenological DNB models assessed in this study. The parametric trends of phenomenological DNB models, however are somewhat abnormal comparing to experimental data. So it can be argued that the existing phenomenological DNB models are rather empirical than theoretical so far

  17. Conceptual design of simplified PWR

    The authors believe the next generation nuclear power plant should be characterized by: (1) simplicity of design; (2) ease of operation and maintenance; (3) economic conformance with safety requirements; and (4) technologies easy to understand by the public. 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, the Japan Atomic Power Company (JAPC) supported by the other Japanese PWR Utilities, Electricite de France (EdF), Westinghouse (WH) and Mitsubishi Heavy Industry (MHI) have studied application of passive technologies at a power rating of about 1,000 MWe. The limited availability for location of nuclear power plant in Japan makes plants with higher power ratings more desirable. Using the AP-600 reference design as a basis, the authors enlarged the plant size to 3 -loops and added engineering features to conform with Japanese practice and Utilities' preference. The Simplified PWR (SPWR) program definitively confirmed the feasibility of a passive plant with an NSSS rating about 1,000 MWe and 3 loops

  18. Safety Review models on radioactive source term design for PWR waste treatment systems

    The source of all liquid, gaseous and solid radioactive waste in the pressure water reactor (PWR) nuclear power plant (NPP) originate from leakage of fission products out of the fuel rods into the primary coolant and neutron activation of materials within and around the primary coolant system and reactor vessel. The source term design used to determine the concentrations of radionuclides in the reactor coolant, which could be: (1) A conservative source term which predicts the maximum concentration of nuclides in the Reactor Coolant System establishes the design basis of the various onsite processing systems for the purpose of defining system capacity and shielding requirements and (2) A realistic source term for the purpose of evaluating the reasonably expected inventories and releases of radionuclides under normal operating condition, including anticipated operational occurrences. This paper will discuss the safety review models on source term design for the PWR waste system mainly on the basis of the conception of the conservative source term. (author)

  19. Crud formation on low duty PWR fuel in the Halden reactor

    A previous paper summarised observations on the effects of water chemistry and thermal-hydraulic conditions on crud formation on PWR fuel in the Halden reactor. These observations led to the conclusion that a critical degree of fuel duty (which can be expressed as degree of coolant sub-cooled boiling, void fraction or mass evaporation rate) was required for the formation of tenacious crud deposits. Recent measurements of the oxide layers on low duty PWR fuel have revealed the formation of tenacious crud deposits. This paper describes the operating history of the fuel rods, including water chemistry and thermal-hydraulic conditions, and suggests reasons for the sudden appearance of the crud deposits. (author)

  20. Shielding and Containment Evaluations of the NAC-LWT Cask with Tritium Burnable Poison Rods

    In 1989, the NAC legal weight truck cask (NAC-LWT) was approved by the U.S. Nuclear Regulatory Commission to transport either one pressurized water reactor (PWR) fuel assembly or two boiling water reactor (BWR) fuel assemblies. Since that time, license amendments have allowed the shipment of high-burnup PWR and BWR fuel rods, MTR-type research reactor fuel elements, and TRIGA-type fuel elements. In 1999, DOE approved an NAC-LWT submittal for a shipment of lead test assemblies (LTAs) containing tritium-producing burnable poison rods (TPBARs). This paper presents the 10 CFR Part 71 shielding and containment evaluations of the NAC-LWT with the LTA payload

  1. A 2D-3D FEM approach of fuel rod thermomechanical behaviour during a RIA

    For better understanding of the fuel rod behaviour during a RIA and to extrapolate the CABRI tests results to PWR conditions, a pellet and its corresponding cladding part have been modelled by means of a 2D axisymmetric meshing, with EDF's finite element code ASTER. The pellet rim region, which is modelled with a 3D meshing, is represented in the global 2D-model with an equivalent homogenized material. The stress distribution is calculated by applying a thermal radial profile computed with the CEA/IPSN SCANAIR code. Then, the local stresses are determined in the rim region, in the neighbourhood of a gas bubble. This 2D-3D FEM approach has been applied successively to REP Na1 rod, at the time and location of the first failure, and to the postulated RCCA ejection accident in a PWR. (R.P.)

  2. Fuel rod failure as a consequence of departure from nucleate boiling or dryout

    PWR and BWR reactor test data on the brittle failure of Zircaloy fuel rod cladding are compared with out-of-pile test data. The reactor test fuel rods were exposed to power-cooling mismatch (PCM) and to consequent departure from nucleate boiling (DNB) or to dryout and consequent clad over-temperature, under PWR and BWR test conditions, respectively. The reactor test data show that cladding integrity is generally maintained despite exposure to very severe accident environments. The cladding time-at-temperature boundaries between the failure and non-failure data from the reactor tests and from the out-of-pile tests are in very good agreement. Therefore, it would appear that brittle-ductile boundary curves generated out-of-pile can be used to predict cladding oxidation embrittlement and subsequent brittle failure which might be caused by reactor upset and accident conditions

  3. Assessment of the influence of subchannel analysis model on the prediction of CHF in rod bundles

    The influence of the turbulent mixing model employed in a subchannel analysis code was investigated in this study, especially on the prediction of the critical heat flux (CHF) in rod bundles. The equal-volume-exchange turbulent mixing and void drift model was employed in the MATRA code, and the void drift coefficient was optimized through the analysis of two-phase flow distribution data for GE 9-rod and Ispra 16-rod test bundles. The influence of the subchannel analysis model on the analysis of CHF was examined by evaluating the CHF test data in rod bundles representing PWR and BWR conditions. The CHFR margin of typical LWR cores was evaluated by taking into account the influence on the local parameter CHF correlation and the hot channel analysis result. As the result, it appeared that the turbulent mixing model has an important effect on the prediction of CHF under the low pressure and the closed-assembly-channel conditions

  4. Development of burnup dependent fuel rod model in COBRA-TF

    Yilmaz, Mine Ozdemir

    The purpose of this research was to develop a burnup dependent fuel thermal conductivity model within Pennsylvania State University, Reactor Dynamics and Fuel Management Group (RDFMG) version of the subchannel thermal-hydraulics code COBRA-TF (CTF). The model takes into account first, the degradation of fuel thermal conductivity with high burnup; and second, the fuel thermal conductivity dependence on the Gadolinium content for both UO2 and MOX fuel rods. The modified Nuclear Fuel Industries (NFI) model for UO2 fuel rods and Duriez/Modified NFI Model for MOX fuel rods were incorporated into CTF and fuel centerline predictions were compared against Halden experimental test data and FRAPCON-3.4 predictions to validate the burnup dependent fuel thermal conductivity model in CTF. Experimental test cases from Halden reactor fuel rods for UO2 fuel rods at Beginning of Life (BOL), through lifetime without Gd2O3 and through lifetime with Gd 2O3 and a MOX fuel rod were simulated with CTF. Since test fuel rod and FRAPCON-3.4 results were based on single rod measurements, CTF was run for a single fuel rod surrounded with a single channel configuration. Input decks for CTF were developed for one fuel rod located at the center of a subchannel (rod-centered subchannel approach). Fuel centerline temperatures predicted by CTF were compared against the measurements from Halden experimental test data and the predictions from FRAPCON-3.4. After implementing the new fuel thermal conductivity model in CTF and validating the model with experimental data, CTF model was applied to steady state and transient calculations. 4x4 PWR fuel bundle configuration from Purdue MOX benchmark was used to apply the new model for steady state and transient calculations. First, one of each high burnup UO2 and MOX fuel rods from 4x4 matrix were selected to carry out single fuel rod calculations and fuel centerline temperatures predicted by CTF/TORT-TD were compared against CTF /TORT-TD /FRAPTRAN

  5. LWR fuel rod testing facilities in high flux reactor (HFT) Petten for investigation of power cycling and ramping behaviour

    LWR fuel rod irradiation experiments are being performed in HFR's Pool Side Facility (PSF) by means of pressurized boiling water capsules (BWFC). For more than 6 years the major application of these devices has been in performing irradiation programs to investigate the power ramp behaviour of PWR and BWR rods which have been pre-irradiated in power reactors. Irradiation devices with various types of monitoring instrumentation are available, e.g. for fuel rod length, fuel stack length, fuel rod internal pressure and fuel rod central temperature measurements. The application scope covers PWR and BWR fuel rods, with burn-up levels up to 45 MWd/kg(U), max. linear heat generation rates up to 700 W/cm and simulation of constant power change rates between 0.05 and 1000 W/cm.min. The paper describes the different designs of LWR fuel rod testing facilities and associated non-destructive testing devices in use at the HFR Petten. It also addresses the new test capabilities that will become available after exchange of the HFR vessel in 1983. Furthermore it shows some typical results. (author)

  6. Device for measuring the diameter of a rod for a nuclear reactor fuel element

    The device is used for the underwater determination of the diameter of a control rod of a PWR over its whole length. The guide is pivoted on the support on two pivots, which are at right angles to one another and to the longitudinal direction of the guide channel. Fitting bodies elastically sprung in the radial direction are provided for the rod on the wall of the guide channel. The second measuring cutter is supported elastically sprung in the direction of the measuring diameter. (orig./HP)

  7. Application of STAV5 code for the analysis of fission gas release in power reactor rods

    STAV5 is a design code for calculation of temperatures, fission gas release and rod pressure in BWR and PWR fuel rods. It includes the treatment of pellet cracks affecting conductivity and thermal expansion, gap closure by eccentric or relocated pellet fragments and oxide and crud build-up on the clad outer surface. The fission gas release model consists of two parts: High temperature release based on grain boundary saturation and low temperature release varying with fission rate of different isotopes. STAV5 has been benchmarked with a number of inpile thermal measurement experiments to as high burnup as 25 MWd/kg U. The main application of STAV5 is as a routine design tool for power reactor rods. It is also used to compare with PIE data. Examples are given from the analyses of fission gas release data from BWR rods from Oskarshamn 1 and Barsebeck 1 as well as PWR rods from Maine Yankee initial cores. The STAV5 evaluation show the importance of power histories, densification and the position in the assembly. (author)

  8. Study of two control rods of a district heating nuclear plant

    This paper broaches the study of the control rods to ensure a convenient working during load following of the nuclear reactor THERMOS. The mathematical model is descriptive of the whole of the nuclear plant (point model for the core and the heat balances). Two power control are studied. The first, like PWR, is a program for the mean temperature of primary water. The second takes into account the structure of the plant and is described by a schedule of powers

  9. Transient performance and design aspects of low boron PWR cores with increased utilization of burnable absorbers

    Papukchiev, Angel [GRS mbH Forschungsinstitute, Garching (Germany); Schaefer, Anselm [ISaR GmbH, Garching (Germany)

    2008-07-01

    In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. As high boron concentrations have significant impact on reactivity feedback properties and core transient behaviour, design changes to reduce boron concentration in the reactor coolant are of general interest in view of improving PWR inherent safety. In order to assess the potential advantages of such strategies in current PWRs, two low boron core configurations based on fuel with increased utilization of gadolinium and erbium burnable absorbers have been developed. The new PWR designs permit to reduce the natural boron concentration in reactor coolant at begin of cycle to 518 (Gd) and 805 (Er) ppm. An innovative low boron core design methodology was implemented combining a simplified reactivity balance search procedure with a core design approach based on detailed 3D diffusion calculations. Fuel cross sections needed for nuclear libraries were generated using the 2D lattice code HELIOS [2] and full core configurations were modelled with the 3D diffusion code QUABOX/CUBBOX [3]. For dynamic 3D calculations, the coupled code system ATHLET - QUABOX/CUBBOX was used [4]. The new cores meet German acceptance criteria regarding stuck rod, departure from nucleate boiling ratio (DNBR), shutdown margin, and maximal linear power. For the assessment of potential safety advantages of the new cores, comparative analyses were performed for three PWR core designs: the already mentioned two low boron designs and a standard design. The improved safety performance of the low boron cores in anticipated transients without scram (ATWS), boron dilution scenarios and beyond design basis accidents (BDBA) has already been reported in [1, 2 and 3]. This paper gives a short reminder on the results obtained. Moreover, it deals not only with the potential advantages, but also addresses the drawbacks of the new PWR configurations - complex core design, increased power

  10. PWR fuel: experience and development

    The start-up of the large French nuclear program has rapidly led FRAGEMA to be one of the first PWR fuel suppliers. FRAGEMA is a joint subsidiary of two companies whose scopes of supply are fully complementary: FRAMATOME (NSSS vendor) and COGEMA (nuclear fuel cycle service supplier). At the center of these two activities FRAGEMA is in charge of designing and marketing fuel assemblies. Assistance is also offered to nuclear power plant operators in all fuel related fields by providing a wide range of services and a number of specialized components. Over the past years a statistical data base has been accumulated on fuel assembly behaviour under various operating conditions. At the same time extensive experimental programs have been, set up to develop advanced products to cope with utilities needs in the future. An overview of these two sides of our experience is presented in the following

  11. PWR standardization: The French experience

    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. Quantitative analysis of gases in fuels. Applications to PWR type reactors fuels

    The different methods used in Saclay and Cadarache to determine the quantity of gases which are present in fuels and fuel cans of PWR type reactors are described. These gases are fission gases (Xe, Kr), pollutant gases (hydrocarbons, N2, O2, H2O, CO, CO2), filling gases (He) or hydrides. A description of the equipment and the operation mode used are given. The obtained results on uranium oxides and mixed oxide fuels are compared with the measures of gases released in the whole rod. (O.M.)

  13. Multi-rod burst test under a loss-of-coolant accident condition, (2)

    Multi-rod burst test No. 7806 was performed with a views to estimating the quantitative channel blockage caused by the ballooning of fuel assembly during a postulated LOCA. The test was conducted under conditions of initial internal pressure 20 kg/cm2, steam flow rate 0.4 g/cm2min and heating rate 90C/sec. Following results were obtained: (1) Internal pressure increased up to 28 kg/cm2 during the heat-up stage, the average burst pressure of 49 rods being measured to be about 26 kg/cm2. (2) Almost all ruptured claddings had relatively short ballooned region length expanded above 34%. The length ranged from 0 to 40 mm, being much shorter than those measured in other tests performed under different conditions. (3) Maximum channel blockage of the assembly(7 x 7) was measured to be 36.2%, while was 43.4% in the interior rods(5 x 5) which had relatively uniform temperature distribution in the radial direction of the rod. These values were also smaller than those measured in other tests. (author)

  14. Sucker rod centralizer

    Rezewski, J.

    1988-01-26

    This patent describes an oil well sucker rod guide consisting of an elongated body having a number of radial slots. Each slot is disposed at equiangular spaced positions, and contains a roller rotatably supported upon an axle transverse to the slot, such that the roller projects outside the periphery of the body from only one end of the slot

  15. Piston for rod pumping

    Pastushenko, G.I.

    1965-06-22

    A piston, or plunger, for rod pumping, is made up of a cylindrical housing with labryinthal seals, a nose piece, and a scraper. In order to remove paraffin from the inside surface of the production pipe, the housing is made in telescopic form. The scraper consists of an arrangement of springs installed on the outer surface of the housing.

  16. Simulation of Spent PWR Fuel Assembly Behavior Under Normal Conditions of Transport

    The behavior of a PWR high-burnup spent fuel assembly under normal conditions of transport is simulated in a dynamic analysis of a 0.3-m free drop of a transportation cask unprotected by impact limiters striking a flat rigid surface in the horizontal orientation. The structural analysis employs a finite element numerical model consisting of the cask, the fuel assemblies, the fuel rods, the guide tubes and the cask’s internal structures that hold the fuel assemblies in position. Appropriate mechanical properties for the cask’s structural components, as well as the elastic-plastic properties typical of high-burnup Zircaloy-4 cladding, are utilized. Emphasis is placed on fuel rods responses at locations where maximum forces would be expected, which include end-plate positions and spacer-grid positions at assembly mid-span. Temporal and spatial variations of the forces acting on the fuel rods are calculated and post-processed to obtain frequency distributions, which statistically represent the total fuel rod population in the cask. The results show that the largest pinch force, (ror-to-rod contact force), is 1700 lb, the maximum axial force is 600 lb, and the largest bending moment is 175 in-lb. Failure analysis of fuel rods using these force quantities, and considering the effects of potential hydrides reorientation on cladding failure resistance, indicates, under conservative assumptions, a factor of safety of least 2 against longitudinal tearing, and no failure is predicted for transverse tearing or rod breakage. Fuel reconfiguration is predicted not to occur, and although partial tearing of guide tubes is possible, it is not enough to impair post-accident assembly retrieval. (author)

  17. Calculation and analysis of heat source of PWR assemblies based on Monte Carlo method

    When fission occurs in nuclear fuel in reactor core, it releases numerous neutron and γ radiation, which takes energy deposition in fuel components and yields many factors such as thermal stressing and radiation damage influencing the safe operation of a reactor. Using the three-dimensional Monte Carlo transport calculation program MCNP and continuous cross-section database based on ENDF/B series to calculate the heat rate of the heat source on reference assemblies of a PWR when loading with 18-month short refueling cycle mode, and get the precise values of the control rod, thimble plug and new burnable poison rod within Gd, so as to provide basis for reactor design and safety verification. (authors)

  18. PWR refill-reflood analysis with experimental loop and calculation model. Pt. 2

    Equations for control volumes varying in the time have been applied. The bottom and length of the bubble and film boiling region in the core are specified by a correlation and time constant based on our measurements. The boiling volume is divided into two parts, saturated water and steam volume. The hydraulic processes are calculated to the average fuel rod, but for the temperatures also the hot rod is calculated. Some parameters have been determined by comparison of measured and calculated results. Sensitivity analyses were made for a PWR, and the hydraulic resistance of the pump (and water stopper evtl. in the loop) was found as the most important factor to ensure a sufficient reflood. (orig.)

  19. Frictional Behavior of Fe-based Cladding Candidates for PWR

    After the recent nuclear disaster at Fukushima Daiichi reactors, there is a growing consensus on the development of new fuel systems (i.e., accident-tolerant fuel, ATF) that have high safety margins under design-basis accident (DBA) and beyond design-basis accident (BDBA). A common objective of various developing candidates is to archive the outstanding corrosion-resistance under severe accidents such as DBA and DBDA conditions for decreasing hydrogen production and increasing coping time to respond to severe accidents. ATF could be defined as new fuel/cladding system with enhanced accident tolerant to loss of active cooling in the core for a considerably longer time period under severe accidents while maintaining or improving the fuel performance during normal operations. This means that, in normal operating conditions, new fuel systems should be applicable to current operating PWRs for suppressing various degradation mechanisms of current fuel assembly without excessive design changes. When considering that one of the major degradation mechanisms of PWR fuel assemblies is a grid-to-rod fretting (GTRF), it is necessary to examine the tribological behavior of various ATF candidates at initial development stage. In this study, friction and reciprocating wear behavior of two kinds of Fe-based ATF candidates were examined with a reciprocating wear tests at room temperature (RT) air and water. The objective is to examine the compatibilities of these Fe-based alloys against current Zr-based alloy properties, which is used as major structural materials of PWR fuel assemblies. The reciprocating wear behaviors of Fe-based accident-tolerant fuel cladding candidates against current Zr-based alloy has been studied using a reciprocating sliding wear tester in room temperature air and water. Frictional behavior and wear depth were used for evaluating the applicability and compatibilities of Fe-based candidates without significant design changes of PWR fuel assemblies

  20. Frictional Behavior of Fe-based Cladding Candidates for PWR

    Lee, Young-Ho; Kim, Hyung-Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Byun, Thak Sang [Oak Ridge National Lab., Oak Ridge (United States)

    2014-10-15

    After the recent nuclear disaster at Fukushima Daiichi reactors, there is a growing consensus on the development of new fuel systems (i.e., accident-tolerant fuel, ATF) that have high safety margins under design-basis accident (DBA) and beyond design-basis accident (BDBA). A common objective of various developing candidates is to archive the outstanding corrosion-resistance under severe accidents such as DBA and DBDA conditions for decreasing hydrogen production and increasing coping time to respond to severe accidents. ATF could be defined as new fuel/cladding system with enhanced accident tolerant to loss of active cooling in the core for a considerably longer time period under severe accidents while maintaining or improving the fuel performance during normal operations. This means that, in normal operating conditions, new fuel systems should be applicable to current operating PWRs for suppressing various degradation mechanisms of current fuel assembly without excessive design changes. When considering that one of the major degradation mechanisms of PWR fuel assemblies is a grid-to-rod fretting (GTRF), it is necessary to examine the tribological behavior of various ATF candidates at initial development stage. In this study, friction and reciprocating wear behavior of two kinds of Fe-based ATF candidates were examined with a reciprocating wear tests at room temperature (RT) air and water. The objective is to examine the compatibilities of these Fe-based alloys against current Zr-based alloy properties, which is used as major structural materials of PWR fuel assemblies. The reciprocating wear behaviors of Fe-based accident-tolerant fuel cladding candidates against current Zr-based alloy has been studied using a reciprocating sliding wear tester in room temperature air and water. Frictional behavior and wear depth were used for evaluating the applicability and compatibilities of Fe-based candidates without significant design changes of PWR fuel assemblies

  1. Control rod cluster with removable rods for nuclear fuel assembly

    For each removable control rod, the open end section of the sleeve has a certain length of reduced diameter with openings in its wall. The top end of the rod is joined to an extension tube that surrounds the shaft over part of its lenght. This extension tube fits over the reduced part of the sleeve when the shaft is screwed into the bore of the sleeve. Rotation of the rod in the sleeve is prevented by deforming the extension tube locally in the openings of the end part of the sleeve. The rod is dismantled by exerting a torque on it using a gripping area near the end of the rod

  2. Morphoelastic rods. Part I: A single growing elastic rod

    Moulton, D.E.

    2013-02-01

    A theory for the dynamics and statics of growing elastic rods is presented. First, a single growing rod is considered and the formalism of three-dimensional multiplicative decomposition of morphoelasticity is used to describe the bulk growth of Kirchhoff elastic rods. Possible constitutive laws for growth are discussed and analysed. Second, a rod constrained or glued to a rigid substrate is considered, with the mismatch between the attachment site and the growing rod inducing stress. This stress can eventually lead to instability, bifurcation, and buckling. © 2012 Elsevier Ltd. All rights reserved.

  3. Method for making sucker rods

    Rasi-Zade, A.T.O.; Kurbanov, N.G.O.; Sutovsky, P.M.; Shikhlinsky, T.M.O.; Kakhramanov, K.T.; Rabinovich, A.M.; Karaev, I.K.O.; Timofeev, V.I.; Ibragimov, O.I.O.

    1991-07-20

    A method for making sucker rods used in oil well pumping is provided, which has the objectives of cutting down the cost of producing sucker rods and of improving their reliability in arduous operating conditions found in wells containing corrosive fluids. The method is characterized in that a rod-body blank is first welded together with rod-end blanks which are made from different materials than the rod-body blank. A welded sucker-rod blank is thus obtained, on which end heads are upset from the blank. The length of the rod-end blank is selected so that weld joints are established, after the upsetting procedure, across the maximum cross-section of the end heads. The method of the invention provides for a weld joint having as much as 3.5 to 4 times the area compared to the rod body, within the zone of the minimum effective stresses acting upon the rod, hence possessing a safety margin of many times the maximum stress applied. This assures high operational reliability and durability of the rods produced according to the invention. The method of the invention does not require precision accuracy in welding the sucker-rod blanks, and minimizes the consumption of expensive alloyed steel, which is used only for making the part of the rod that is subjected to the greatest loads. 7 figs.

  4. Development of an advanced 16x165 Westinghouse type PWR fuel assembly for Slovenia

    Industrias Nucleares do Brasil (INB), KEPCO Nuclear Fuel Company, Ltd. (KNFC), and Westinghouse Electric Company (Westinghouse) have jointly designed an advanced 16x16 Westinghouse type PWR fuel assembly. This advanced 16x16 Westinghouse type PWR fuel assembly, which will be implemented in both Kori Unit 2 (in Korea) and Angra Unit 1 (in Brazil) in January and March 2005, respectively, is an integral part of the utilities fuel management strategy. This same fuel design has also been developed for future use in Krsko Unit 1 (in Slovenia). In this paper we will describe the front-end nuclear fuel management activities utilized by the joint development team and describe how these activities played an integral part in defining the direction of the advanced 16x16 Westinghouse type PWR fuel assembly design. Additionally, this paper will describe how this design demonstrates improved margins under high duty plant operating conditions. The major reason for initiating this joint development program was to update the current 16x16 fuel assembly, which is also called 16STD. The current 16STD fuel assembly contains a non-optimized fuel rod diameter for the fuel rod pitch (i.e. 9.5 mm OD fuel rods at a 0.485 inch pitch), non-neutronic efficient components (i.e. Inconel Mid grids), no Intermediate Flow Mixer (IFM) grids, and other mechanical features. The advanced 16x16 fuel assembly is being designed for peak rod average burnups of up to 75 MWd/kgU and will use an optimized fuel rod diameter (i.e. 9.14 mm OD ZIRLOTM fuel rods), neutronic efficient components (i.e. ZIRLOTM Mid grids), ZIRLOTM Intermediate Flow Mixer (IFM) grids to improve Departure from Nucleate Boiling (DNB) margin, and many other mechanical features that improve design margins. Nuclear design activities in the areas of fuel cycle cost and fuel management were performed in parallel to the fuel assembly design efforts. As the change in reactivity due to the change in the fuel rod diameter influences directly the

  5. The BG18, a B(U)F type package used for the transport of irradiated fuel rods - return of experience

    The purpose of this presentation is to share the return of experience of Transnubel after a period of nearly 3 years operation of the BG18 package in several nuclear power plants and hot cell facilities. This package has been used mainly for the shipment of full scale as well as samples of irradiated fuel rods - UOX or MOX, PWR or BWR

  6. The BG18, a B(U)F type package used for the transport of irradiated fuel rods - return of experience

    Juergen, S.; Herman, S. [Transnubel, Dessel (Belgium)

    2004-07-01

    The purpose of this presentation is to share the return of experience of Transnubel after a period of nearly 3 years operation of the BG18 package in several nuclear power plants and hot cell facilities. This package has been used mainly for the shipment of full scale as well as samples of irradiated fuel rods - UOX or MOX, PWR or BWR.

  7. Dynamic characteristics and design criteria of fuel rod assemblies in a baffle jet flow

    During recent refuelling operations of a PWR type nuclear power plant it was found that several fuel assemblies located at baffle joint were damaged. It was assumed that the damage had been caused by severe vibration of fuel rods induced by coolant leakage through baffle joint. Model testing was con-ducted to identify the vibration mechanism and to obtain the safety criteria for fuel assemblies in a baffle jet flow. Fuel rods are long beams supported along their length by seven grid assemblies. Those prototype rods were simulated as single span simply supported beams. Model assemblies are 4 x 4 and 5 x 4 bundles of simply supported beams with a pitch ratio of 1.33. Flow tests were carried out in a water loop of 40 GPM. It was found that rod assemblies in a jet flow experience large amplitude vibration caused by jet induced instability. The stability boundary of rod assemblies is determined to be Vc/fD=2.3 √(D/h)(msub(o) deltasub(o))/(rho D2). Based on the stability boundary provided, safety limit of baffle gap is calculated as to be 1.6 x 10-3 in. The effect of the position of fuel rod assemblies relative to the baffle joint was investigated. And it was found that the susceptibility of rod assemblities to vibration increases as the stand off distance shortens. (Author)

  8. Evaluation of fuel centerline temperature of LWR fuel rods during first rise to operating power

    In order to reveal principal mechanism which is dominating LWR fuel rod failure due to pellet-cladding interaction, in-reactor irradiation experiments at Halden Boiling Water Reactor (HBWR), Norway have been undertaken by Japan Atomic Energy Research Institute. The test titled as ''Halden Power Ramp Test'' has been initiated since 1979. In the test, fifteen Japanese fuel rods which have equivalent specification of 17 x 17 PWR rods and 8 x 8 BWR rods of current commercial power plants are involved. The fuel rods for future power ramping are now in base irradiation stage under the simulated condition such as typical light water reactor. In spite of their base irradiation stage, it is possible to get in-reactor and on-power data by means of instrumented equipments to measure fuel centerline temperature as well as to measure rod plenum pressure as a function of linear heat generating rate. Analytical evaluation of these data at the beginning-of-irradiation was performed here concentrating on fuel thermal behavior under the operation. Experimental variables in the analysis were addressed to two fabricating parameters such as diametral gap and pure helium fill gas pressurization. The results of evaluation are described in detail. It is the specific feature of this study to include experimental facts and related analyses that enable us to understand thermal behavior of fuel rods under the operating condition of commercial power plants. (author)

  9. Control rod calibration including the rod coupling effect

    In a reactor containing more than one control rod, which includes all reactors licensed in the United States, there will be a 'coupling' or 'shadowing' of control rod flux at the location of a control rod as a result of the flux depression caused by another control rod. It was decided to investigate this phenomenon further, and eventually to put calibration table data or formulae in a small computer in the control room, so once could insert the positions of the three control rods and receive the excess reactivity without referring to separate tables. For this to be accomplished, a 'three control- rod reactivity function' would be used which would include the flux coupling between the rods. The function is design and measured data was fitted into it to determine the calibration constants. The input data for fitting the trial functions consisted of 254 data points, each consisting of the position of the reg, shim, and transient rods, and the total excess reactivity. (About 200 of these points were 'critical balance points', that is the rod positions for which reactor was critical, and the remainder were determined by positive period measurements.) Although this may be unrealistic from a physical viewpoint, the function derived gave a very accurate recalculation of the input data, and thus would faithfully give the excess reactivity for any possible combination of the locations of the three control rods. The next step, incorporation of the three-rod function into the minicomputer, will be pursued in the summer and fall of 1984

  10. IFPE/US-PWR-16 X 16 Lead Test Assembly Extended Burnup Demonstration Program

    Description: US-PWR 16 x 16 LTA (lead test assembly) extended burnup demonstration program conducted during the 1980's. Relevant program data was obtained from the project final report and other supporting documents. The objective of this program was to demonstrate improved nuclear fuel utilization through more efficient fuel management and increased discharge burnup. The use of the 16 x 16 LTAs with Zr-4 cladding in this program demonstrated the capability to achieve peak fuel rod average burnups of ∼ 60 GWd/MTU. Both pool side (non-destructive) and hot cell (destructive) post irradiation examinations (PIE) of selected rods from the two LTAs were conducted. These examinations included rods irradiated for 3 and 5 cycles. Pool side examinations of the LTAs included visual inspection, dimensional measurements, eddy currant testing (ECT), and waterside corrosion thickness measurement. Hot cell fuel rod PIE included void volume measurements, fill gas analyses, cladding visual inspections, dimensional measurements, neutron radiography, and gamma scanning. Fuel pellet examinations included fuel densification and swelling measurements, fuel burnup analyses, and ceramography. Cladding examinations included metallography, hydrogen concentration measurement, and mechanical property testing. The irradiation of two 16 x 16 LTAs was completed in a US commercial PWR. LTA D039 was irradiated during reactor cycles 2 through 4. The irradiation of LTA D040 was extended through reactor cycle 6 to achieve a lead rod, axial average burnup of 58 GWd/MTU. The fuel assembly design consisted of 236 rods in a 16 x 16 array, five control element guide tubes, 12 fuel rod spacer grids, upper and lower end fittings, and a hold-down device. The bottom spacer grid is Inconel 625. All other spacer grids and all guide tubes are Zr-4. The standard fuel rod design consists of enriched UO2, solid cylindrical pellets, a round wire Type 302 stainless steel compression spring, and an alumina spacer

  11. Maintenance robot for PWR plant

    The remote operation, automatic machines utilized in the field of the maintenance of component machinery and equipment in nuclear power plants, so-called maintenance robots, have produced effects in the reduction of radiation exposure, the improvement of the quality of working, the shortening of working time and so on, but still many robots have their specialized functions. The expectation of present day society to robots has been diversified, and the technical development of high function robots is advanced positively. In this report, the recent examples of the high function robots developed for PWR power stations with the support of technical progress and the trend of the technical development are explained. The needs and seeds of maintenance robot development are discussed. As the examples of heightening the functions of maintenance robots, the next generation ultrasonic testing machine highly advanced by sensor technology and size and weight reduction mechanism technology, the intelligent monitoring system for welding using AI technology and other manpower-saving robots are shown. (K.I.)

  12. Control rod drive

    A control device for long time stopping is disposed for stopping by applying the same number of normal rotation and reverse rotation to a ball spindle at a predetermined time interval. Even in a case where a control rod is not operated for a long period of time, sticking between a sealing material and a ball spindle is prevented, rotational torque is not increased excessively, and the control rod can always be operated normally. Further, a stopping control device is disposed for applying rotation after the stop of the rotation of the ball spindle in the reverse direction within one turn. Lubricants and obstacles are introduced between the surfaces purified by the rotation, to prevent the direct contact of the purified surfaces to each other and correct the deformation of sealing members. Therefore, the rotational torque is not increased excessively. (N.H.)

  13. Sucker rod centralizer

    Rivas, O.; Newski, A.

    1989-10-03

    This patent describes a device for centralizing at least one sucker rod within a production pipe downhole in a well and for reducing frictional forces between the pipe and at least one sucker rod. It comprises an elongate, substantially cylindrical body member having a longitudinal axis, a plurality of slots within the member and a rotatable member mounted within each slot, each of the plurality of slots has its major dimension along a first axis parallel to the longitudinal axis of the body member and is oriented with respect to the other seats so as to form a helicoidal array for maximizing the total surface contact area between the rotatable members and the pipe and for decreasing the forces acting on each rotatable member.

  14. Fuel rod attachment system

    A reusable system for removably attaching a nuclear reactor fuel rod to a support member. A locking cap is secured to the fuel rod and a locking strip is fastened to the support member or vice versa. The locking cap has two opposing fingers and shaped to form a socket having a body portion. The locking strip has an extension shaped to rigidly attach to the socket's body portion. The locking cap's fingers are resiliently deflectable. For attachment, the locking cap is longitudinally pushed onto the locking strip causing the extension to temporarily deflect open the fingers to engage the socket's body portion. For removal, the process is reversed. In an alternative embodiment, the cap is rigid and the strip is transversely resiliently compressible. (author)

  15. Control rod position detector

    The device of the present invention can save blowers for compulsory cooling. That is, the control rod position detector comprises (1) a control rod driving shaft made of a ferromagnetic material moving in a pressure vessel of a nuclear reactor and (2) detector coils arranged to the outside of the pressure vessel each at an identical distance over the moving stroke of the driving shaft for detecting the position of the driving shaft by the change of inductance. In addition, heat insulation materials are disposed between the detector coils and the reactor pressure vessel. Then, heat from the reactor pressure vessel can be insulated. Accordingly, temperature of the detector coils can be reduced by natural cooling. As a result, since it is no more necessary to dispose compulsory cooling fans as required in a conventional case, the entire device can be constituted economically, and the reliability of the device is improved. (I.S.)

  16. Segment fuel rods

    Purpose: To maintain the integrity of segment fuel rods without causing power spikes in adjacent fuel rods. Constitution: Power spikes are generated in the portions in adjacent with end plugs of segment fuel elements shielded welding zircaloy end plugs, because water/uranium ratio is locally increased due to the absence of pellets at the bonding end plugs to increase the neutron moderating effect thereby increasing the thermal neutron fluxes to rise the reactivity. This can be prevented most effectively by absorbing excess neutron absorbers. In view of the above, the purpose can be attained either by incorporating high neutron absorbing material at the bonding end plug, or constituting the bonding end plug itself with neutron absorbing material. (Kamimura, M.)

  17. Fuel rod fixing system

    This is a reusable system for fixing a nuclear reactor fuel rod to a support. An interlock cap is fixed to the fuel rod and an interlock strip is fixed to the support. The interlock cap has two opposed fingers, which are shaped so that a base is formed with a body part. The interlock strip has an extension, which is shaped so that this is rigidly fixed to the body part of the base. The fingers of the interlock cap are elastic in bending. To fix it, the interlock cap is pushed longitudinally on to the interlock strip, which causes the extension to bend the fingers open in order to engage with the body part of the base. To remove it, the procedure is reversed. (orig.)

  18. The OSCAR code package: a unique tool for simulating PWR contamination

    Understanding the PWR primary circuit contamination by corrosion products, fission products and actinides is a crucial issue for reactor operation and design. The main challenges are decreasing the impact on personnel exposure to radiation, optimizing the plant operation, limiting the activity of the wastes produced during the reactor lifetime and preparing decommissioning. In cooperation with EDF and AREVA NP, CEA has developed the OSCAR code package, a unique tool for simulating PWR contamination. The OSCAR package results from the merging of two codes, which simulate PWR contamination by fission products and actinides (PROFIP code) and by activated corrosion products (PACTOLE code). These two codes have been validated separately against an extensive set of data obtained over 40 years from in-situ gamma spectrometry measurements, sampling and analysing campaigns of primary coolant, as well as experiments in test loops or experimental reactors, which are representative of PWR conditions. In this paper, a new step is presented with the OSCAR code package, combining the features of the two codes and motivated by the fact that, wherever they originate from, the contamination products are subject to the same severe conditions (300 °C, 150 bar, neutron flux, water velocity up to 15 m.s-1) and follow the same transport mechanisms in the primary circuit. The main processes involved are erosion/deposition, dissolution/precipitation, adsorption/desorption, convection, purification, neutron activation, radioactive decrease. The V1.1 version of the OSCAR package is qualified for fission products (Xe, Kr, I, Sr), actinides (U, Np, Pu, Am, Cm) and corrosion products (Ni, Fe, Co, Cr). This paper presents the different release modes (defective fuel rod release, fissile material dissemination, material corrosion and release), then the processes which govern contamination transfer, and finally, we give examples of the comparison of the OSCAR package results with measurements in

  19. Characterization and modeling of the thermal hydraulic and chemical environment of fuel claddings of PWR reactors during boiling

    In pressurised water reactors (PWR), nucleate boiling can strongly influence the oxidation rate of the fuel cladding. To improve our understanding of the effect of the boiling phenomenon on corrosion kinetics, information about the chemical and thermal hydraulic boundary conditions at the heating rod surface is needed. Moreover, very few data are available in the range of thermal hydraulic parameters of PWR cores (15,5 MPa and 340 deg C) concerning the two-phase flow pattern close to the fuel cladding. A visualization device has been adapted on an out-of-pile loop Reggae to obtain both qualitative and quantitative data. These observations provide a direct access to the geometrical properties of the vapor inclusions, the onset of nucleate boiling and the gas velocity and trajectory. An image processing method has been validated to measure both void fraction and interfacial area concentration in a bubbly two-phase flow. Thus, the visualization device proves to be a suitable and accurate instrumentation to characterize nucleate boiling in PWR conditions. The experimental results analysis indicates that a local approach is needed for the modelling of the fuel rod chemical environment. To simulate the chemical additives enrichment, a new model is proposed where the vapor bubbles are now considered as physical obstacles for the liquid access to the rod surface. The influence of the two-phase flow pattern appears to be of major importance for the enrichment phenomenon. This study clearly demonstrates the existence of strong interactions between the two-phase flow pattern, the rod surface condition, the corrosion process and the water chemistry. (author)

  20. Sucker rod guide

    White, R.C.

    1988-10-25

    This patent describes an improved guide for use in a string of sucker rods for reciprocation in a tubing string in a borehole, the sucker rods having threaded male ends, the guide comprising: an elongated upright cylindrical member of external diameter less than the internal diameter of tubing in which it is to be used, the member having sucker rod receiving female threaded openings at the upper and lower ends, the threaded openings being coaxial of the member cylindrical axis whereby the member may be positioned in a string of sucker rods, and including a plurality of spaced-apart parallel sided slots within the member, each slot being of semi-circular configuration and of depth greater than the radius and less than the diameter of the cylindrical member, the sidewalls of each slot being parallel to and equally spaced from a plane of the member cylindrical axis; the member having an axle bore therein for each of the slots, the axle bores being parallel and spaced apart from each other, a plane of the axis of each bore being perpendicular the member cylindrical axis and the axis of each bore being displaced away from the member cylindrical axis; an axle received in each axle bore; and a wheel received on each axle the diameter of each wheel being approximately the diameter of the cylindrical member, the periphery of each wheel extending beyond the member cylindrical wall whereby the wheels are positioned to engage and roll on the internal cylindrical surface of tubing, the planes of adjacent slots in the member being rotationally displaced from each other, a portion of each wheel extending beyond the cylindrical surface of the member, the opposed portion of each wheel being within the confines of the member cylindrical surface whereby each wheel can contact a tubing wall at only one point on its cylindrical surface.

  1. Behavior of irradiated PWR fuel under simulated RIA conditions. Results of the NSRR tests GK-1 and GK-2

    Results from power burst tests, GK-1 and GK-2, conducted at the Nuclear Safety Research Reactor (NSRR), are summarized. The objectives of the tests are to investigate irradiated pressurized water reactor (PWR) fuel behaviors under reactivity initiated accident (RIA) conditions. The tests were performed on a 14 x 14 PWR fuel rod irradiated to a burnup of 42.1 MWd/kgU in the Genkai unit no.1 of Kyushu Electric Power Co., Inc. Test method, data from pre- and post-pulse fuel examinations, transient records during the pulse-irradiations are described and discussed. GK-1 and -2 test fuel rods are short-sized rods re-fabricated from a full size fuel rod. The instrumented test fuel rod in a double-container-type capsule was subjected to the pulse-irradiation with stagnant water cooling condition at 0.1 MPa and 293 K. Deposited energy and peak fuel enthalpy were 505 J/g·fuel and 389 J/g·fuel in the Test GK-1, and 490 J/g·fuel and 377 J/g·fuel in the Test GK-2, respectively. During the pulse-irradiations, departure from nucleate boiling (DNB) occurred and the cladding surface temperature reached 581 K and 569 K in the Tests GK-1 and -2, respectively. The maximum cladding hoop strain was 2.7% in the Test GK-1 and 1.2% in the Test GK-2. However, the test fuel rods did not fail. Estimated fission gas releases during the pulse-irradiations were 11.7% and 7.0% in the Tests GK-1 and -2, respectively. (author)

  2. Method for making sucker rods

    Karaev, I.K.O.; Shikhlinsky, T.M.O.; Polikhronov, K.P.; Sutovsky, P.M.; Avakian, E.V.; Semkin, N.V.; Rabinovich, A.M.; Dzhabarov, R.D.

    1991-01-15

    A method for making sucker rods composed of a rod body and end heads is provided. The rod body end portions are subjected to an upsetting procedure which is carried out at a temperature that precludes softening of the rod body metal. A thickening is formed on each of the end portions, whose width in a direction square with the rod body axis is equal to or exceeds the head maximum diameter in the place of the weld joint, and whose length exceeds the width of the heat-affected zone involved in the welding process. A transition portion is shaped as a solid of revolution whose cross-section smoothly and continuously decreases from the thickening towards the rod body. The upsetting procedure is followed by pressure welding of each of the end heads together with the thickening on the rod body end portion and by turning the weld joint zone.

  3. Sucker rod coupling

    Klyne, A.A.

    1986-11-11

    An anti-friction sucker rod coupling is described for connecting a pair of sucker rods and centralizing them in a tubing string, comprising: an elongate, rigid, substantially cylindrical body member, each end of the body member forming means for threadably connecting the body member with a sucker rod. The body member further forms a transversely extending, substantially diametric, generally vertical slot extending therethrough. The body member further forms a pin bore, such pin bore extending transversely through the body member so as to intersect the slot substantially perpendicularly; a wheel member positioned within the slot to rotate in a generally vertical plane. The wheel member has a portion thereof extending beyond the periphery of the body member to engage the inner surface of the tubing string and centralize the coupling; and a pin mounted in the pin bore and supporting member thereon, whereby the wheel member is rotatable within the slot; the wheel member having sufficient clearance between its side surfaces and the wall surfaces of the slot, when the wheel member is centered in the slot on the pin, whereby the wheel member may shift along the pin to assist in ejecting sand and oil from the slot.

  4. Control rod drive mechanisms

    Purpose: To accurately measure the loads generated upon scram and judge the absence or presence of deceleration in control rod drive mechanisms. Constitution: Control rod drive mechanisms for use in a BWR type reactor includes an index tube vertically movably, connected at the upper end to the control rod and having a drive piston at the lower end. A piezoelectric member for detecting the load generated upon uprise of the index tube is disposed and signals from the piezoelectric member is connected to a calculation processing device. A load exerted when the index tube uprises is measured by way of the piezoelectric member upon scram thereby judging the absence or presence of the decelerating operation. Therefore, the nuclear reactor can be shutdown only when it is required with no excess safety operation than required. As a result, the reactor availability can be improved and, in addition, it is also possible to mitigate the burden of in-service inspection and reduce the operators' exposure. (Kamimura, M.)

  5. Nuclear fuel rods

    Purpose: To enable a tight seal in fuel rods while keeping the sealing gas pressure at an exact predetermined pressure in fuel rods. Constitution: A vent aperture and a valve are provided to the upper end plug of a cladding tube. At first, the valve is opened to fill gas at a predetermined pressure in the fuel can. Then, a conical valve body is closely fitted to a valve seat by the rotation of a needle valve to eliminate the gap in the engaging thread portion and close the vent aperture. After conducting the reduced pressure test for the fuel rod in a water tank, welding joints are formed between the valve and the end plug through welding to completely seal the cladding tube. Since the welding is conducted after the can has been closed by the valve, the predetermined gas pressure can be maintained at an exact level with no efforts from welding heat and with effective gas leak prevention by the double sealing. (Kawakami, Y.)

  6. Nuclear criticality safety analysis for the traveller PWR fuel shipping package

    The Traveller PWR fresh fuel shipping package represents a radical departure from conventional PWR fuel package designs. Two immediately noticeable features of the Traveller are that it carries a single fuel assembly instead of two as do other package designs, and that it has built-in moderator, which forms part of the flux-trap system. The criticality safety case shows that the Traveller satisfies both U.S. and IAEA licensing requirements, and demonstrates that the package remains acceptably subcritical under normal conditions and hypothetical accident conditions of transport. This paper looks at the modeling techniques that were used to analyze the several accident scenarios that were considered, including: Lattice pitch expansion; Lattice pitch expansion along the fuel assembly length; Preferential flooding (selective flooding of different cavities); Differential flooding (varying water levels inside different cavities); Partial flooding (varying water density); Axial rod displacement; o Sensitivity studies of variable foam densities and boron content in packaging; Analysis for carrying loose rods in a rodbox; The criticality safety case for the Traveller proved to be a successful cooperative effort between ENUSA and Westinghouse

  7. Nupec thermal hydraulic test to evaluate post-DNB characteristics for PWR fuel assemblies (1. general test plan and results)

    In the present thermal hydraulic design of Pressurized Water Reactor (PWR), a departure from nucleate boiling (DNB) under anticipated transient conditions is not allowed. However, it is recognized that the DNB dose not cause a fuel rod failure immediately, and a suitable reactor trip can prevent the core from severe damages. If the fuel rod temperature under the post-DNB conditions can be accurately evaluated, the potentially existing margin in the present design method will be quantitatively assessed. To establish the heat transfer evaluation method on post-DNB event for PWR thermal hydraulic design, Nuclear Power Engineering Corporation (NUPEC) started a program, NUPEC Thermal Hydraulic Test to Evaluate Post-DNB Characteristics for PWR Fuel Assemblies (NUPEC-TH-P), in 1995 (hereinafter the year means fiscal year) under the sponsorship of Ministry of Economy, Trade and industry (METI). This program is now under going until 2001. This paper is to show the overall plan and the status of NUPEC-TH-P. (authors)

  8. AP1000® PWR reactor physics analysis with VERA-CS and KENO-VI. Part 2. Power distribution

    Westinghouse has applied the Core Simulator of the Virtual Environment for Reactor Applications, VERA-CS, under development by the Consortium for Advanced Simulation of LWRs (CASL) to the core physics analysis of the AP1000® PWR. The AP1000 PWR features an advanced first core with radial and axial heterogeneities, including enrichment zoning, multiple burnable absorbers, and a combination of light and heavy control banks to enable the MSHIMTM advanced operational strategy. These advanced features make application of VERA-CS to the AP1000 PWR first core especially relevant to qualify VERA performance. A companion paper at this conference describes the results obtained with VERA-CS and the KENO Monte-Carlo code for startup physics tests simulations of the AP1000 PWR first core (critical boron, rod worth and reactivity coefficients). This paper describes the results of detailed power distribution comparisons between VERA-CS and KENO, and confirms the excellent numerical agreement reported in the companion paper for the startup physics tests simulations. (author)

  9. Analysis of reactivity worths of highly-burnt PWR fuel samples measured in LWR-PROTEUS Phase II

    The reactivity loss of PWR fuel with burnup has been determined experimentally by inserting fresh and highly-burnt fuel samples in a PWR test lattice in the framework of the LWR-PROTEUS Phase II programme. Seven UO2 samples irradiated in a Swiss PWR plant with burnups ranging from ∼40 to ∼120 MWd/kg and four MOX samples with burnups up to ∼70 MWd/kg were oscillated in a test region constituted of actual PWR UO2 fuel rods in the centre of the PROTEUS zero-power experimental facility. The measurements were analyzed using the CASMO-4E fuel assembly code and a cross section library based on the ENDF/B-VI evaluation. The results show close proximity between calculated and measured reactivity effects and no trend for a deterioration of the quality of the prediction at high burnup. The analysis thus demonstrates the high accuracy of the calculation of the reactivity of highly-burnt fuel. (authors)

  10. Thermodynamic modelling of PWR coolant

    Corrosion products released from PWR and VVER primary circuit surface oxides are transported in the coolant to the core, where they deposit and are activated to form radioactive corrosion products, which can be re-released to re-deposit on out-of-core surfaces. Spinel solubilities vary with the pH, temperature and sometimes the hydrogen concentration of the coolant. This paper describes the development of an equilibrium thermodynamic model to predict such changes, and discusses the extent of the available solubility data for Fe, Ni, Co and Zn oxides. Results are described on the relative solubility of Fe and Ni under both normal operating conditions and during shutdown/start-up, and on the relative stabilities of stoichiometric and non-stoichiometric zinc ferrite spinels. Comparison of the calculated corrosion product concentrations with reactor measurements indicates that, in reactors with low Ni content in the steam generator alloys, the concentration of Ni in the coolant is limited by its availability in the surface oxide. In reactors with high-Ni alloys, the circulating Ni concentrations may be dominated by colloidal material. The calculated changes in Ni and Fe concentrations during the acid-reducing phase of shutdown are in reasonable agreement with measurements from Sizewell B. The paper highlights the need for a more comprehensive open corrosion product data base, the need to consider both boiling and radiolysis in the core on corrosion product solubility in different parts of the primary circuit and, finally, the importance of kinetic factors at low temperature behaviour during shutdown and start-up. (author)

  11. EB welding and quality control of nuclear reactor fuel rods at ASEA-ATOM

    Fourteen years ago ASEA-ATOM chose EB welding for fuel rod plug/tube welds. This choice was made on the basis of 7 years of experience of EB-welding of fuel rods in a pilot plant. The specific reasons were the high quality and the high process yield, which are made possible by the great degree of controlability and reproducibility of this process and because the welds are suitable for QC inspection by an inline ultrasonic method which we developed at the same time. To date ASEA-ATOM has manufactured approximately 600,000 fuel rods with 1,200,000 EB-welds. The results have met expections as regards quality, process yield and service in BWR and PWR reactors. Descriptions are given of the automatic Sciaky EB welding machines, of the ultrasonic inspection equipment and of their process qualification. Some comments are made on quality and process yield

  12. Advanced PWR fuel design concepts

    For nearly 15 years, Combustion Engineering has provided pressurized water reactor fuel with the features most suppliers are now introducing in their advanced fuel designs. Zircaloy grids, removable upper end fittings, large fission gas plenum, high burnup, integral burnable poisons and sophisticated analytical methods are all features of C-E standard fuel which have been well proven by reactor performance. C-E's next generation fuel for pressurized water reactors features 24-month operating cycles, optimal lattice burnable poisons, increased resistance to common industry fuel rod failure mechanisms, and hardware and methodology for operating margin improvements. Application of these various improvements offer continued improvement in fuel cycle economics, plant operation and maintenance. (author)

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

    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)

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

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

  15. Activity transport models for PWR primary circuits

    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.)

  16. Program of monitoring PWR fuel in Spain

    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)

  17. Assessment of spectral history influence on PWR and WWER core

    The few-group cross section libraries, used by reactor dynamics codes, are affected by the spectral history effect - a dependence of fuel cross sections not only on burnup, but also on local spectral conditions during burnup. Neglecting this effect leads to an additional component of error in neutron-physical characteristics. Two solution approaches to this problem implemented in the reactor dynamic code DYN3D are described and compared in this paper: a cross section correction method based on 239Pu concentration and separate cross sections treatment for each axial layer of reactor core. Steady-state and burnup characteristics of a PWR and a WWER-1000 cores, calculated by DYN3D with and without cross section corrections, are compared. An impact of the correction on transient calculations is studied for a control rod ejection example. Studies have shown a significant influence of spectral history on axial power and burnup distributions as well as on calculated cycle length. Two different correction methods have shown similar major effects. (orig.)

  18. Effects of generation and optimization of libraries of effective sections in the analysis of transient in PWR reactors; Efectos de generacion y optimizacion de librerias de secciones eficaces en el analisis de transitorios en reactores PWR

    Sanchez-Cervera, S.; Garcia Herranz, N.; Cuervo, D.; Ahnert, C.

    2014-07-01

    In this paper evaluates the impact that has a certain mesh on a transient in a PWR reactor in the expulsion of a control bar. Have been used for this purpose the coupled codes neutronic and Thermo-hydraulic COBAYA3/COBRA-TF. This objective has been chosen the OECD/NEA PWR MOX/UO{sub 2} rod ejection transient benchmark provides isotopic compositions and defined geometric configurations that allow the use of codes lattice to generate own bookstores. The code used for this transport has been the code APOLLO2.8. The results show large discrepancies when using the benchmark library or libraries own by comparing them to the other participants solutions. The source of these discrepancies is the nodal effective sections provided in the benchmark. (Author)

  19. Pre-pulse irradiation examination, NSRR pulse irradiation and post-pulse irradiation examination of MH-1 fuel rod

    The Nuclear Safety Research Reactor (NSRR) program for studying failure threshold of pre-irradiated LWR fuel under simulated reactivity initiated accident conditions is in progress. In this program a 14 x 14 PWR type fuel K4-1 was segmented from K4/G08 long size PWR rod pre-irradiated in MIHAMA Unit-2 and was pulse irradiated on November 28, 1989 at NSRR. Energy deposition given to the test rod was 60 cal/g·fuel. No failure indication was observed by in-core monitoring and by post-pulse irradiation examination. As one of the NSRR data base on fuel behavior during transient/RIA, data obtained from pre-pulse irradiation examination, during NSRR pulse irradiation, and from post-pulse irradiation examination are summarized. (author)

  20. Pre-pulse irradiation examination, NSRR pulse irradiation and post-pulse irradiation examination of MH-2 fuel rod

    The Nuclear Safety Research Reactor (NSRR) program for studying failure threshold of pre-irradiated LWR fuel under simulated reactivity initiated accident conditions is in progress. In this program, a 14 x 14 PWR type fuel K4-2 was segmented from a K4/G08 long size PWR rod pre-irradiated in MIHAMA Unit-2 and was pulse irradiated on March 8, 1990 at NSRR. Energy deposition given to the test rod was 68 cal/g·fuel. No failure indication was observed by in-core monitoring and by post-pulse irradiation examination. As one of the NSRR data base on fuel behavior during transient/RIA, data obtained from pre-pulse irradiation examination, during NSRR pulse irradiation, and from post-pulse irradiation examination are summarized. (author)

  1. Development and design of control rod drive mechanisms for pressurized water reactors

    The Control Rod Drive Mechanisms (CRDM) for a Pressurized Water Reactor (PWR) are equipment, integrated to the reactor pressure vessel, incorporating mechanical and electrical components designed to move and position the control rods to guarantee the control of power and shutdown of the nuclear reactor, during normal operation, either in emergency or accidental situations. The type of CRDM used in PWR reactors, whose detailed individual description will be presented in this monograph are the Roller-Nut and Magnetic-Jack. The environment, where the CRDM performs its above presented operational functions, includes direct contact with the fluid used as coolant peculiar to the interior of the reactor, and its associated chemical characteristics, the radiation field next to the reactor core, and also the temperature and pressure in the reactor pressure vessel. So the importance of the CRDM design requirements related to its safety functions are emphasized. Finally, some aspects related to the mechanical and structural design of CRDM of a case study, considering the CRDM for a PWR from the experimental nuclear plant to be applied by CTMSP (Centro Tecnologico da Marinha em Sao Paulo), are pointed out. The design and development of these equipment (author)

  2. The THYC three-dimensional thermal-hydraulic codes for rod bundles: Recent developments and tests

    Pressurized water reactor (PWR) or liquid-metal fast breeder reactor cores or fuel assemblies, PWR steam generators, condensers, and tubular heat exchangers are basic components of a nuclear power plant that involve two-phase flows in tube or rod bundles. A deep knowledge of the detailed flow patterns on the shell side is necessary to evaluate departure from nucleate boiling (DNB) margins in reactor cores, singularity effects (grids, wire spacers, support plates, and baffles), corrosion on the steam generator tube sheet, bypass effects, and vibration risks. For that purpose, Electricite de France has developed since 1986 a general purpose thermal-HYdraulic Code (THYC) to study three-dimensional single- and two-phase flows in rod or tube bundles (PWR codes, steam generators, condensers, and heat exchangers). It considers the three-dimensional domain to contain two kinds of components: fluid and solids. The THYC model is obtained by space-time averaging of the instantaneous equations (mass, momentum, and energy) of each phase over control volumes including fluid and solids. The physical model of THYC is validated under several French and international experiments for single- and two-phase flows. The THYC is used for the calculation of transients such as steam-line break (coupled with a three-dimensional neutronics code), for DNB predictions, and for various steam generator or condenser studies

  3. CRC DEPLETION CALCULATIONS FOR THE RODDED ASSEMBLIES IN BATCHES 1, 2, 3, AND 1X OF CRYSTAL RIVER UNIT 3

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain rodded fuel assemblies from batches 1, 2, 3, and 1X of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A rodded assembly is one that contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) for some period of time during its irradiation history. The objective of this analysis is to provide SAS2H calculated isotopic compositions of depleted fuel and depleted burnable poison for each fuel assembly to be used in subsequent CRC reactivity calculations containing the fuel assemblies

  4. CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 1, 2, AND 3 OF CRYSTAL RIVER UNIT 3

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 1, 2, and 3 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies

  5. CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 1, 2, AND 3 OF CRYSTAL RIVER UNIT 3

    Kenneth D. Wright

    1997-07-29

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 1, 2, and 3 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies.

  6. CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 4 AND 5 OF CRYSTAL RIVER UNIT 3

    Kenneth D. Wright

    1997-07-30

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 4 and 5 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies.

  7. CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 4 AND 5 OF CRYSTAL RIVER UNIT 3

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 4 and 5 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies

  8. Turbulent flows in complex rod bundle geometries numerically predicted by the use of FEM and a basic turbulence model

    Newer projects in nuclear reactor design tend to higher conversion ratios. An up-to-date PWR has a conversion ratio of approximately 0.53. Whereas APWRs are planned to have up to 0.95 and the breeder reactor is supposed to have a ratio better than 1.0. High conversion ratios necessitate tightly packed fuel rod lattices. Together with high burnup, necessary for economic efficiency, the slender fuel rods show a tendency to bend. The result of bent fuel rods is a distorted lattice. A further item which leads to irregular lattices are the tolerances in fuel rod assembly. The absolute values of tolerance which can be seen as fixed become relatively more important in tightly packed lattices. (orig./GL)

  9. CRC DEPLETION CALCULATIONS FOR THE RODDED ASSEMBLIES IN BATCHES 1, 2, 3, AND 1X OF CRYSTAL RIVER UNIT 3

    Kenneth D. Wright

    1997-09-03

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain rodded fuel assemblies from batches 1, 2, 3, and 1X of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A rodded assembly is one that contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) for some period of time during its irradiation history. The objective of this analysis is to provide SAS2H calculated isotopic compositions of depleted fuel and depleted burnable poison for each fuel assembly to be used in subsequent CRC reactivity calculations containing the fuel assemblies.

  10. CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 8 AND 9 CRYSTAL RIVER UNIT 3

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 8 and 9 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies

  11. Large-eddy simulations of turbulent flow for grid-to-rod fretting in nuclear reactors

    Bakosi, J; Lowrie, R B; Pritchett-Sheats, L A; Nourgaliev, R R

    2013-01-01

    The grid-to-rod fretting (GTRF) problem in pressurized water reactors is a flow-induced vibration problem that results in wear and failure of the fuel rods in nuclear assemblies. In order to understand the fluid dynamics of GTRF and to build an archival database of turbulence statistics for various configurations, implicit large-eddy simulations of time-dependent single-phase turbulent flow have been performed in 3x3 and 5x5 rod bundles with a single grid spacer. To assess the computational mesh and resolution requirements, a method for quantitative assessment of unstructured meshes with no-slip walls is described. The calculations have been carried out using Hydra-TH, a thermal-hydraulics code developed at Los Alamos for the Consortium for Advanced Simulation of Light water reactors, a United States Department of Energy Innovation Hub. Hydra-TH uses a second-order implicit incremental projection method to solve the single-phase incompressible Navier-Stokes equations. The simulations explicitly resolve the la...

  12. Aerosol behavior during SIC control rod failure in QUENCH-13 test

    In a nuclear reactor severe accident, radioactive fission products as well as structural materials are released from the core by evaporation, and the released gases form particles by nucleation and condensation. In addition, aerosol particles may be generated by droplet formation and fragmentation of the core. In pressurized water reactors (PWR), a commonly used control rod material is silver-indium-cadmium (SIC) covered with stainless steel cladding. The control rod elements, Cd, In and Ag, have relatively low melting temperatures, and especially Cd has also a very low boiling point. Control rods are likely to fail early on in the accident due to melting of the stainless steel cladding which can be accelerated by eutectic interaction between stainless steel and the surrounding Zircaloy guide tube. The release of the control rod materials would follow the cladding failure thus affecting aerosol source term as well as fuel rod degradation. The QUENCH experimental program at Forschungszentrum Karlsruhe investigates phenomena associated with reflood of a degrading core under postulated severe accident conditions. QUENCH-13 test was the first in this program to include a silver-indium-cadmium control rod of prototypic PWR design. To characterize the extent of aerosol release during the control rod failure, aerosol particle size distribution and concentration measurements in the off-gas pipe of the QUENCH facility were carried out. For the first time, it was possible to determine on-line the aerosol concentration and size distribution released from the core. These results are of prime importance for model development for the proper calculation of the source term resulting from control rod failure. The on-line measurement showed that the main aerosol release started at the bundle temperature maximum of T ∼ 1570 K at hottest bundle elevation. A very large burst of aerosols was detected 660 s later at the bundle temperature maximum of T ∼ 1650 K, followed by a

  13. Fiber optic laser rod

    Erickson, G.F.

    1988-04-13

    A laser rod is formed from a plurality of optical fibers, each forming an individual laser. Synchronization of the individual fiber lasers is obtained by evanescent wave coupling between adjacent optical fiber cores. The fiber cores are dye-doped and spaced at a distance appropriate for evanescent wave coupling at the wavelength of the selected dye. An interstitial material having an index of refraction lower than that of the fiber core provides the optical isolation for effective lasing action while maintaining the cores at the appropriate coupling distance. 2 figs.

  14. Cone rod dystrophies

    Hamel Christian P

    2007-02-01

    Full Text Available Abstract Cone rod dystrophies (CRDs (prevalence 1/40,000 are inherited retinal dystrophies that belong to the group of pigmentary retinopathies. CRDs are characterized by retinal pigment deposits visible on fundus examination, predominantly localized to the macular region. In contrast to typical retinitis pigmentosa (RP, also called the rod cone dystrophies (RCDs resulting from the primary loss in rod photoreceptors and later followed by the secondary loss in cone photoreceptors, CRDs reflect the opposite sequence of events. CRD is characterized by primary cone involvement, or, sometimes, by concomitant loss of both cones and rods that explains the predominant symptoms of CRDs: decreased visual acuity, color vision defects, photoaversion and decreased sensitivity in the central visual field, later followed by progressive loss in peripheral vision and night blindness. The clinical course of CRDs is generally more severe and rapid than that of RCDs, leading to earlier legal blindness and disability. At end stage, however, CRDs do not differ from RCDs. CRDs are most frequently non syndromic, but they may also be part of several syndromes, such as Bardet Biedl syndrome and Spinocerebellar Ataxia Type 7 (SCA7. Non syndromic CRDs are genetically heterogeneous (ten cloned genes and three loci have been identified so far. The four major causative genes involved in the pathogenesis of CRDs are ABCA4 (which causes Stargardt disease and also 30 to 60% of autosomal recessive CRDs, CRX and GUCY2D (which are responsible for many reported cases of autosomal dominant CRDs, and RPGR (which causes about 2/3 of X-linked RP and also an undetermined percentage of X-linked CRDs. It is likely that highly deleterious mutations in genes that otherwise cause RP or macular dystrophy may also lead to CRDs. The diagnosis of CRDs is based on clinical history, fundus examination and electroretinogram. Molecular diagnosis can be made for some genes, genetic counseling is

  15. Technical description of the NRC long-term whole-rod and crud performance test

    Einziger, R.E.; Fish, R.L.; Knecht, R.L.

    1982-09-01

    Westinghouse Hanford Company (WHC) and EG and G-Idaho are jointly conducting a long-term, low-temperature, spent-fuel, whole rod and crud behavior test to provide the Nuclear Regulatory Commission (NRC) with information to assist in the licensing of light water reactor (LWR) spent-fuel, dry storage facilities. Readily available fuel rods from an H.B. Robinson Unit 2 (PWR) fuel assembly and a Peach Bottom-II (BWR) fuel assembly were selected for use in the 50-month test. Both intact and defected rods will be tested in inert and oxidizing atmospheres. A 230/sup 0/C test temperature was selected for the first 10-month run. Both nondestructive and destructive examinations are planned to characterize the fuel rod behavior during the 5-y test. Four interim examinations and a final examination will be conducted. Crud spallation behavior will be investigated by sampling the crud particulate from the test capsules at each of the four interim examinations and at the end of the test. The background to whole rod testing, description of rod breach mechanisms, and a detailed description of the test are presented in this document.

  16. Technical description of the NRC long-term whole-rod and crud performance test

    Westinghouse Hanford Company (WHC) and EG and G-Idaho are jointly conducting a long-term, low-temperature, spent-fuel, whole rod and crud behavior test to provide the Nuclear Regulatory Commission (NRC) with information to assist in the licensing of light water reactor (LWR) spent-fuel, dry storage facilities. Readily available fuel rods from an H.B. Robinson Unit 2 (PWR) fuel assembly and a Peach Bottom-II (BWR) fuel assembly were selected for use in the 50-month test. Both intact and defected rods will be tested in inert and oxidizing atmospheres. A 2300C test temperature was selected for the first 10-month run. Both nondestructive and destructive examinations are planned to characterize the fuel rod behavior during the 5-y test. Four interim examinations and a final examination will be conducted. Crud spallation behavior will be investigated by sampling the crud particulate from the test capsules at each of the four interim examinations and at the end of the test. The background to whole rod testing, description of rod breach mechanisms, and a detailed description of the test are presented in this document

  17. Design of an on-line detection system for fuel rod failure in a pressurized water reactor

    An on-line detection system for fuel rod failure of pressurized water reactor was designed to give a realtime indication of the activity concentrations of key radionuclides in the primary coolant. Tests were performed for the system in a PWR and the real-time monitoring data of the primary coolant was obtained. The tests showed that the system is sensitive to the activity concentrations of key radionuclides in the primary coolant and is able to detect the fuel rod failure. Minimum detectable concentrations for key radionuclides 135Xe and 88Kr are of the order of 3 and 10 kBq L-1, respectively. (author)

  18. Improved fuel rod support means

    A fuel bundle for a nuclear reactor having a plurality of fuel rods supported between spaced tie plates, wherein coolant flows through said tie plates and past said fuel rods, characterized by: an end plug disposed between an end of each fuel rod and the adjacent tie plate, and means defining a passage for the flow of coolant through the interface between said end plug and said tie plates to minimize crud buildup at said interface

  19. RAPTA-5 code: Modelling behaviour of WWER-type fuel rods in design basis accidents verification calculations

    RAPTA-5 code used for licensing calculations to validate the compliance with the requirements for WWER fuel safety in design basis accidents. The characteristic results are given of design modelling experiments simulating thermomechanical and corrosion behaviour of WWER and PWR fuel rods in LOCA. The results corroborate the adequate predictability of both individual design models and the code as a whole. (author). 14 refs, 12 figs

  20. Piston and connecting rod assembly

    Brogdon, James William (Inventor); Gill, David Keith (Inventor); Chatten, John K. (Inventor)

    2001-01-01

    A piston and connecting rod assembly includes a piston crown, a piston skirt, a connecting rod, and a bearing insert. The piston skirt is a component separate from the piston crown and is connected to the piston crown to provide a piston body. The bearing insert is a component separate from the piston crown and the piston skirt and is fixedly disposed within the piston body. A bearing surface of a connecting rod contacts the bearing insert to thereby movably associate the connecting rod and the piston body.

  1. Metrology of irradiated fuel rods

    A nuclear fuel rod has its diameter measured by back illumination from a lamp which casts a silhouette image of the rod on to a photo diode array. A radiation shielding wall exists between the rod and array with optical transmission through the wall. The output of the array is threshold voltage level sensed to generate a data pulse which envelops clock pulses digitally representing the rod diameter. The resolution is improved by interpolating between diode position in the array by means of an integrator. (author)

  2. Development of CHF correlation “MG-NV” for low pressure and low velocity conditions applied to PWR safety analysis

    The Critical Heat Flux (CHF) is one of the important parameters in the safety analysis of Pressurized Water Reactor (PWR). If the CHF is reached, an abrupt drop occurs in the heat transfer between the fuel rod cladding and the reactor coolant, which may induce a large temperature excursion of fuel cladding and a subsequent fuel failure. Therefore, accurate prediction of CHF is required in order to assure a sufficient safety margin in the PWR core. Mitsubishi Heavy Industries, ltd (MHI) is developing a new series of CHF correlations which covers various fuel designs and wide range of fluid conditions with sufficient reliability. In this paper, a new CHF correlation, MG-NV (Mitsubishi Generalized correlation for Non-Vane grid spacers) is presented. This correlation is one of the basic components of the new correlation series and was developed to cover low pressure and low velocity conditions where the rod bundle CHF data are limited. The CHF correlation was developed based on open CHF database and provides conservative but more reliable rod bundle CHF predictions compared with the conventional CHF correlations used in safety analyses at low pressure condition, such as Main Steam Line Break event. (author)

  3. Fuel rod plugs

    Purpose: To prevent the formation of voids to the inside of welded portion in fuel rod plugs. Constitution: A fuel rod is tightly sealed by welding end plugs at both ends of a fuel can charged with nuclear fuel material. For the welding of the end plug, laser welding has now been employed with the reason of increasing the welding efficiency and reducing the welding heat distortion. However, if the end plug is laser-welded to the end of the fuel can in the conventional form, there is a problem that voids are liable to be formed near the deepest penetration in the welding portion. That is, gases evolved near the deepest penetration remains in a key-hole like welded metal portion to result in voids there. Accordingly, grooves capable of passing the laser beam key hole therethrough are disposed along the circumferential direction of the pipe at the end plug welded portion in the fuel can. In this way, since gases generating near the deepest penetration are discharged into the grooves, the key hole-like welded metal is completely filled and voids are not formed. (Kamimura, M.)

  4. Control rod drive system

    The present invention concerns an electromotive driving-type control rod driving system of a BWR type reactor, for which sliding resistance (friction) test can be performed of a movable portion of the control rod driving mechanisms. Namely, a hydraulic pressure control unit has following constitutions in addition to a conventional constitution as a sliding resistance test performing function. (1) A restricting valve is disposed downstream of the scram valve of scram pipelines to control flow rate and pressure of pressurized water flown in the pipelines. (2) A pressure gauge detects a pressure between the scram valve and the restricting valve. (3) A flow meter detects the flow rate of pipelines controlled by the restricting valve. (4) A recording pressure detector detects the pressure at the downstream of the restricting valve. (5) The recording device is attached when the sliding resistant test is performed for tracing the pressure measured by the pressure detection device. Further, the scram valve sends electric signals to a central operation chamber when it is fully closed. The central operation chamber has a function of fully opening the restricting valve by way of the electric signals. (I.S.)

  5. Study of power peak migration due to insertion of control bars in a PWR reactor

    This paper aims to present a study on the power distribution behavior in a PWR reactor, considering the intensity and the migration of power peaks as is the insertion of control rods in the core banks. For this, the study of the diffusion of neutrons in the reactor was adopted by computer simulation that uses the finite difference method for numerically solving the neutron diffusion equation to two energy groups in steady state and in symmetry of a fourth quarter core. We decided to add the EPRI-9R 3D benchmark thermal-hydraulic parameters of a typical power PWR. With a new configuration for the reactor, the positions of the control rods banks were also modified. Due to the new positioning of these banks in the reactor, there was intense power gradients, favoring the occurrence of critical situations and logically unconventional for operation of a nuclear reactor. However, these facts have led interesting times for the study on the power distribution behavior in the reactor, showing axial migration of power peaks and mainly the effect of the geometry of the core on the latter. Based on the distribution of power was evident the increase of the power in elements located in the central region of the reactor core and, concomitantly, the reduction in elements of its periphery. Of course, the behavior exhibited by the simulated reactor is not in agreement with that expected in an actual reactor, where the insertion of control rods banks should lead to reduced power throughout the core as evenly as possible, avoiding sharp power peaks, standardizing the burning fuel, controlling reactivity deviations and acting in reactor shutdown

  6. Simulation model of a PWR power plant

    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)

  7. PWR reactors for BBR nuclear power plants

    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)

  8. Full MOX core design for PWR

    Full MOX core design for APWR was analyzed in nuclear design, fuel integrity analysis, thermal hydraulic design and safety analysis et. al. Feasibility of Full MOX core was confirmed from these analyses without any large modifications. Full MOX PWR core has very good characteristics in which single Pu content in an assembly, burnable poison free, higher burnup and longer cycle operation are feasible. (author)

  9. An evaluation of tight - pitch PWR cores

    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.)

  10. AP1000® PWR reactor physics analysis with VERA-CS and KENO-VI. Part 1. Zero power physics tests

    Westinghouse has applied the Core Simulator of the Virtual Environment for Reactor Applications, VERA-CS, under development by the Consortium for Advanced Simulation of LWRs (CASL) to the core physics analysis of the AP1000® PWR. The AP1000 PWR features and advanced first core with radial and axial heterogeneities, including enrichment zoning, multiple burnable absorbers, and a combination of light and heavy control banks to enable the MSHIMTM advanced operational strategy. These advanced features make application of VERA-CS to the AP1000 PWR first core especially relevant to qualify VERA performance. A companion paper at this conference describes the power distribution analysis of the AP1000 PWR with VERA-CS and the KENO Monte-Carlo code. This paper describes the results obtained for the startup physics tests simulations of the AP1000 PWR first core (critical boron, rod worth and reactivity coefficients), supporting the excellent numerical agreement reported in the companion paper for the power distribution. (author)

  11. The effect of power ramping on fuel rod behaviour

    The behaviour under irradiation of a typical proposed small P.W.R. (100 MWth) fuel rod has been studied with the COMETHE Code-Code for predicting mechanical and thermal fuel performance the fuel rod has been assumed to be irradiated in low power rated core zone during three cycles of operation. The COMETHE results at the end of cycle two (E0C2) have been saved in order to study the power ramping effect on the fuel rod behaviour stress induced in the clad-and to optimize the best way of reaching the full power. Also, in the design of fuel pins, a precise knowledge of the thermal and mechanical behaviour of the rods during irradiation time is required. This involves a realistic evaluation of a number of interrelated parameters, e.g., the temperature distribution, the reconstructing and swelling rates of fuel pellets, the dimensions, the stresses and the strains in the clad, the inner gas pressure, etc. This complex problem can only be properly handed by a computer programme which analyses the fuel pin thermal and mechanical behaviour at successive steps of its irradiation life. COMETHE III-J is a multipurpose code which can provide the user with a realistic solution of this complex problem. The COMETHE III-J has been tested and calibrated with experimental results and here after, there will come some of applications of this code: 1. Evaluation of fuel vendor proposal. 2. Comparison between different proposed designs. 3. Analysis of the effects of tolerances. 4. Analysis of the effects of power history, power ramping, core management, effects anticipated by vendor and operation limitations imposed by the vendor, etc. 5. Help for selection of PIE and interpretation of the results. 6. Analysis of steady state input values for LOCA analysis. 7. Improvement of plant availability and manoeuvrability through reduction of pin failure probability and minimization of limitations from LOCA. (Author)

  12. A systematic approach for development of a PWR cladding corrosion model

    A new model for the in-reactor corrosion of Improved (low-tin) Zircaloy-4 cladding irradiated in commercial pressurized water reactors (PWRs) is described. The model is based on an extensive database of PWR fuel cladding corrosion data from fuel irradiated in commercial reactors, with a range of fuel duty and coolant chemistry control strategies which bracket current PWR fuel management practices. The fuel thermal duty with these current fuel management practices is characterized by a significant amount of sub-cooled nucleate boiling (SNB) during the fuel's residence in-core, and the cladding corrosion model is very sensitive to the coolant heat transfer models used to calculate the coolant temperature at the oxide surface. The systematic approach to developing the new corrosion model therefore began with a review and evaluation of several alternative models for the forced convection and SNB coolant heat transfer. The heat transfer literature is not sufficient to determine which of these heat transfer models is most appropriate for PWR fuel rod operating conditions, and the selection of the coolant heat transfer model used in the new cladding corrosion model has been coupled with a statistical analysis of the in-reactor corrosion enhancement factors and their impact on obtaining the best fit to the cladding corrosion data. The in-reactor corrosion enhancement factors considered in this statistical analysis are based on a review of the current literature for PWR cladding corrosion phenomenology and models. Fuel operating condition factors which this literature review indicated could have a significant effect on the cladding corrosion performance were also evaluated in detail in developing the corrosion model. An iterative least squares fitting procedure was used to obtain the model coefficients and select the coolant heat transfer models and in-reactor corrosion enhancement factors. This statistical procedure was completed with an exhaustive analysis of the model

  13. Nuclear reactor with control rods

    The invention relates to liquid cooled nuclear reactors. In particular, it concerns reactors with mobile control rods in a straight line and guide tubes to guide these control rods through the internal upper components of the reactor vessel and in the aligned fuel assemblies of the core

  14. Tunable frequency 4-rod RFQ

    The frequency tunability of the 4-rod RFQ is investigated. By moving the shorting plate between the posts, which support the 4-rod electrodes, the resonant frequency can be varied almost twofold. Model studies and the calculation of the field uniformity, the variation of Q, and the shunt impedance are reported. (author)

  15. Sucker rod pump

    Brewer, J.R.

    1992-04-14

    This patent describes a subsurface well pump, it comprises: a working barrel; a plunger which reciprocates along the vertical axis within the working barrel between an upper and lower position; a rod connected to the plunger and extending to a means for providing reciprocating force; a well string extending from the top of the working barrel to the surface; an outlet check valve which permits flow to exit the working barrel into the well string and does not permit flow to exit the well string into the working barrel; and an inlet check valve which permits flow into the working barrel from outside of the subsurface pump, the inlet check valve being above the top position of the plunger, the inlet check valve having a cross sectional flow area about equal to or greater than the horizontal cross sectional area of the working barrel, and the inlet check valve being a hinged flapper valve.

  16. Fuel rod reprocessing plant

    A plant for the reprocessing of fuel rods for a nuclear reactor comprises a plurality of rectangular compartments desirably arranged on a rectangular grid. Signal lines, power lines, pipes, conduits for instrumentation, and other communication lines leave a compartment just below its top edges. A vehicle access zone permits overhead and/or mobile cranes to remove covers from compartments. The number of compartments is at least 25% greater than the number of compartments used in the initial design and operation of the plant. Vacant compartments are available in which replacement apparatus can be constructed. At the time of the replacement of a unit, the piping and conduits are altered to utilize the substitute equipment in the formerly vacant compartment, and it is put on stream prior to dismantling old equipment from the previous compartment. Thus the downtime for the reprocessing plant for such a changeover is less than in a traditional reprocessing plant

  17. Control rod drives

    Purpose: To improve the reliability of a device for driving an LMFBR type reactor control rod by providing a buffer unit having a stationary electromagnetic coil and a movable electromagnetic coil in the device to thereby avord impact stress at scram time and to simplify the structure of the buffer unit. Constitution: A non-contact type buffer unit is constructed with a stationary electromagnetic coil, a cable for the stationary coil, a movable electromagnetic coil, a spring cable for the movable coil, and a backup coil spring or the like. Force produced at scram time is delivered without impact by the attracting or repelling force between the stationary coil and the movable coil of the buffer unit. Accordingly, since the buffer unit is of a non-contact type, there is no mechanical impact and thus no large impact stress, and as it has simple configuration, the reliability is improved and the maintenance can be conducted more easily. (Yoshihara, H.)

  18. Nuclear fuel rod

    Purpose: To prevent eutectic reaction between coil spring material and end plug material at the welding work of fuel fabrication. Constitution: Close-contact windings are formed at the end of a coil spring, and base end of a stainless steel supporting member is screwed to the close-contact winding portion of the coil spring. The other end of the supporting member is formed in a conical shape whose apex is in contact with the center of the bottom surface of a zirconium alloy end plug of a cladding tube. In the fuel rod thus constructed, the heating temperature of the end contact portion of the supporting member, at the time of welding the end plug to the cladding tube, can be somewhat lower than the eutectic temperatures of iron, chromium, nickel (the main ingredients of the stainless steel) and zirconium (the main ingredient of the end plug), and accourdingly no eutectic reaction occurs. (Yoshihara, H.)

  19. Acoustic sensor for in-pile fuel rod fission gas release measurement

    We have developed a specific acoustic sensor to improve the knowledge of fission gas release in Pressurized Water Reactor (PWR) fuel rods when irradiated in materials testing reactors. In order to perform experimental programs related to the study of the fission gas release kinetics, the CEA (French Nuclear Energy Commission) acquired the ability to equip a pre-irradiated PWR fuel rod with three sensors, allowing the simultaneous on-line measurements of the following parameters: - fuel temperature with a centre-line thermocouple type C, - internal pressure with a specific counter-pressure sensor, - fraction of fission gas released in the fuel rod with an innovative acoustic sensor. The third detector is the subject of this paper. This original acoustic sensor has been designed to measure the molar mass and pressure of the gas contained in the fuel rod plenum. For in-pile instrumentation, the fraction of fission gas, such as Krypton and Xenon, in Helium, can be deduced online from this measurement. The principle of this acoustical sensor is the following: a piezoelectric transducer generates acoustic waves in a cavity connected to the fuel rod plenum. The acoustic waves are propagated and reflected in this cavity and then detected by the transducer. The data processing of the signal gives the velocity of the acoustic waves and their amplitude, which can be related respectively to the molar mass and to the pressure of the gas. The piezoelectric material of this sensor has been qualified in nuclear conditions (gamma and neutron radiations). The complete sensor has also been specifically designed to be implemented in materials testing reactors conditions. For this purpose some technical points have been studied in details: - fixing of the piezoelectric sample in a reliable way with a suitable signal transmission, - size of the gas cavity to avoid any perturbation of the acoustic waves, - miniaturization of the sensor because of narrow in-pile experimental devices

  20. Research on Operation and Control Strategy of 600MW PWR in Load Follow

    600MW Pressurized Water Reactor (PWR) is designed to operate in Constant Axial Offset Control (CAOC) strategy with base load originally. By calculations over a typical load follow scenario '12-3-6-3 (100-50-100%FP) via the CASMO-4E and SIMULATE-3 package, values of core operating parameter have been examined. With the progress of the nuclear power industry, advanced reactors are considered to have a good performance in load follow, economy and flexibility. Under the premise of fuel loading and structural dimensions unchanged, two independent control rod groups M and AO are used in 600MW pressurized water reactor to provide fine control of both the core reactivity and axial power distribution, which is named ' Improved G strategy .' The influences of different control rod distributions, composition materials, and overlap steps had in power changes have been examined in a comparative study to choose the optimal one.Then we simulate a range of load follow scenarios of the redesigned 600MW core without adjusting soluble boron concentration in the begin, middle and end of first cycle. This paper additionally demonstrated the moderator temperature coefficient and shutdown margin values of the reactor in Improved G strategy to compare with the thermal safety design criteria. It's demonstrated that adequate adjustment of control rod groups enable the core to perform load follow through Improved G strategy in 80% of cycle and save a large volume of liquid effluent particularly toward the end of cycle

  1. A cold stepping driving experiment for a control rod drive line of PWR

    A fatigue analysis for high stress location of a drive shaft is underteken under the impact load when the shaft is driven step by step. The amplitude analysis program of 7T17S signal processor and M.A. Miner's linear cumulative damage theory are used

  2. Microstructural characterization and properties of dissimilar joints used in coupling of PWR control rod driving

    The chemical, mechanical and microstructural characterizations of a dissimilar joint between SA336F347 austenitic and SA479Tp414 martensitic stainless steels were done, welded by TIG process, defining as a result of this characterization that the ER Ni Cr-3 Ni consumable seems to be the best applicable consumable compared to the ER309L consumable; The main variables of the process control were also evaluated, its weldability and properties for a future qualification of a welding procedure, besides to simulate possible situations to be found in this type of joint, such as, its weldability by the LASER process, welded joint without filler metal and without shielding gas, obtaining in this way enough data for the production of products that contains this type of joint. (author)

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

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

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

  4. Physics Analysis of a Prismatic VHTR with Asymmetric Control Rods by Using the HELIOS/MASTER Code Package

    A new physics analysis procedure is under development for prismatic VHTRs based on a conventional two-step procedure for a PWR physics analysis. The HELIOS and MASTER codes were employed to generate the coarse group cross sections through a transport lattice calculation, and to perform the 3-dimensional core physics analysis by a nodal diffusion calculation, respectively. Since prismatic VHTRs such as a GT-MHR include asymmetrically located large control rods, a control rod treatment is a challenging issue in a physics analysis. Previously, we performed a physics analysis for a prismatic VHTR in which symmetric control rods were assumed. Large spectrum shifts due to a control rod insertion on the surrounding blocks could be covered by optimizing the coarse energy group structure. However, it was noted that some improvements should be made in the prediction of the reaction rates and the control rod worths. In this study a new analysis procedure has been developed to deal with asymmetric control rods more accurately. Surface dependent discontinuity factors obtained from multi-block models were applied to the core calculations for a better prediction of the reaction rates and control rod worths. Benchmark calculations were performed for the GT-MHR cores, where the reference solutions were obtained from the MCNP calculations

  5. Status of rod consolidation, 1988

    It is estimated that the spent fuel storage pools at some domestic light-water reactors will run out of space before 2003, the year that the US Department of Energy currently predicts it will have a repository available. Of the methods being studied to alleviate the problem, rod consolidation is one of the leading candidates for achieving more efficient use of existing space in spent fuel storage pools. Rod consolidation involves mechanically removing all the fuel rods from the fuel assembly hardware (i.e., the structural components) and placing the fuel rods in a close-packed array in a canister without space grids. A typical goal of rod consolidation systems is to insert the fuel rods from two fuel assemblies into a canister that has the same exterior dimensions as one standard fuel assembly (i.e., to achieve a consolidation or compaction ratio of 2:1) and to compact the nonfuel-bearing structural components from those two fuel assemblies by a factor of 10 to 20. This report provides an overview of the current status of rod consolidation in the United States and a small amount of information on related activities in other countries. 85 refs., 36 figs., 5 tabs

  6. A Multi-Stage Wear Model for Grid-to-Rod Fretting of Nuclear Fuel Rods

    Blau, Peter Julian [ORNL

    2014-01-01

    The wear of fuel rod cladding against the supporting structures in the cores of pressurized water nuclear reactors (PWRs) is an important and potentially costly tribological issue. Grid-to-rod fretting (GTRF), as it is known, involves not only time-varying contact conditions, but also elevated temperatures, flowing hot water, aqueous tribo-corrosion, and the embrittling effects of neutron fluences. The multi-stage, closed-form analytical model described in this paper relies on published out-of-reactor wear and corrosion data and a set of simplifying assumptions to portray the conversion of frictional work into wear depth. The cladding material of interest is a zirconium-based alloy called Zircaloy-4, and the grid support is made of a harder and more wear-resistant material. Focus is on the wear of the cladding. The model involves an incubation stage, a surface oxide wear stage, and a base alloy wear stage. The wear coefficient, which is a measure of the efficiency of conversion of frictional work into wear damage, can change to reflect the evolving metallurgical condition of the alloy. Wear coefficients for Zircaloy-4 and for a polyphase zirconia layer were back-calculated for a range of times required to wear to a critical depth. Inputs for the model, like the friction coefficient, are taken from the tribology literature in lieu of in-reactor tribological data. Concepts of classical fretting were used as a basis, but are modified to enable the model to accommodate the complexities of the PWR environment. Factors like grid spring relaxation, pre-oxidation of the cladding, multiple oxide phases, gap formation, impact, and hydrogen embrittlement are part of the problem definition but uncertainties in their relative roles limits the ability to validate the model. Sample calculations of wear depth versus time in the cladding illustrate how GTRF wear might occur in a discontinuous fashion during months-long reactor operating cycles. A means to account for grid/rod gaps

  7. Horizontal Drop of 21- PWR Waste Package

    The objective of this calculation is to determine the structural response of the waste package (WP) dropped horizontally from a specified height. The WP used for that purpose is the 21-Pressurized Water Reactor (PWR) WP. The scope of this document is limited to reporting the calculation results in-terms of stress intensities. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 16). AP-3.12Q, ''Calculations'' (Ref. 1 1) is used to perform the calculation and develop the document. The sketches attached to this calculation provide the potential dimensions and materials for the 21-PWR WP design

  8. Thorium fuel cycle study for PWR applications

    A nuclear design feasibility of thorium fueled high converting PWR was investigated. Two kinds of fuel design option were tested for the comparison with conventional UO2 fuel. The first one was an application of MHTGR pyro-carbon coated particle fuels. The other design was an application of MOX fuels as a ThO2-PuO2 ceramic pellet. In the case of carbon-coated particle fuels, there was no benefit in nuclear design aspect because enrichment of U-235 was required over 5 w/o in order to match with the K-infinite of Ulchin-3/4 fuels. However, the use of thorium based plutonium fuels in PWR gave favorable aspects in nuclear design such as flatter K-infinite curve, lower M. T. C. and lower F. T. C. than that of UO2 fuel. (author). 6 refs., 3 tabs., 6 figs

  9. Horizontal Drop of 21- PWR Waste Package

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

  10. Thorium fuel cycle study for PWR applications

    Lim, Jae Yong; Kim, Myung Hyun [Kyung Hee Univ., Seoul (Korea, Republic of)

    1997-12-31

    A nuclear design feasibility of thorium fueled high converting PWR was investigated. Two kinds of fuel design option were tested for the comparison with conventional UO{sub 2} fuel. The first one was an application of MHTGR pyro-carbon coated particle fuels. The other design was an application of MOX fuels as a ThO{sub 2}-PuO{sub 2} ceramic pellet. In the case of carbon-coated particle fuels, there was no benefit in nuclear design aspect because enrichment of U-235 was required over 5 w/o in order to match with the K-infinite of Ulchin-3/4 fuels. However, the use of thorium based plutonium fuels in PWR gave favorable aspects in nuclear design such as flatter K-infinite curve, lower M. T. C. and lower F. T. C. than that of UO{sub 2} fuel. (author). 6 refs., 3 tabs., 6 figs.

  11. Single rod leak detection and repair of leaking or damaged fuel assemblies

    In some circumstances, it is necessary to perform rework operations on some fuel assemblies in order to make them reusable in reactors, movable, transportable or consistent with fuel reprocessor specifications, depending on the plant utility policy. These rework operations are of two types: - Those which consist in restoring the leak tightness of the fuel assemblies. They are made after a series of tests allowing the localization of the failed fuel rods: at first, overall leak detection is provided by monitoring primary coolant activity during reactor operation; then, during refuelling, leaking assemblies are identified by subjecting each of the assemblies scheduled for reloading to a sipping test; finally individual leaking fuel rods are singled out before the defective assemblies can be repaired, i.e. failed rods can be replaced. - Those which involve replacement of part of or the whole assembly structure (combined or not with replacement of failed fuel rods). In order to meet these two needs for rework operations, FRAGEMA has developed a full range of test and tooling systems for detecting single leaking rods in irradiated fuel assemblies and for restoring fuel assemblies to be used in PWR nuclear power plants. As an illustration of the means available, two of these systems are described

  12. Space-dependent dynamics of PWR

    The azimuthal dependent reactor dynamics coupled to thermohydraulics are studied by using the neutron-flux and coolant temperature signals measured at an actual PWR. The second azimuthal mode of neutron-flux fluctuation was found, and the coupling of the mode to thermohydraulics of the coolant was suggested. The coherent coolant flow in the reactor core seems to sustain this spatial oscillation mode. (authors)

  13. Sensitivity analysis of a PWR pressurizer

    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)

  14. PWR fuel behavior: lessons learned from LOFT

    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

  15. Optimum fuel use in PWR reactors

    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.)

  16. Chemical and radiochemical specifications - PWR power plants

    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.)

  17. Steam shut-off valves for PWR type reactors

    Fast acting closure means are requested in PWR type reactors as well as in BWR to safely shut-off the live steam at the turbine input in the event of accident. The design and control system of steam shut-off valves acted by the fluid system and intended for PWR type reactors, are described. The role of these valves in a PWR is discussed with the specified requirements involved

  18. Modeling the activity of 129I and 137Cs in the primary coolant and CVCS resin of an operating PWR

    Mathematical models have been developed to describe the activities of 129I and 137Cs in the primary coolant and resin of the chemical and volume control system (CVCS) during constant power operation in a pressurized water reactor (PWR). The models, which account for the source releases from defective fuel rod(s) and tramp uranium, rely on the contribution of CVCS resin and boron recovery system as a removal process, and differences in behavior for each nuclide. The current models were validated through measured coolant activities of 137Cs. The resultant scaling factors agree reasonably well with the results of the test resin of the coolant and the actual resins from the PWRs of other countries

  19. Project description: ORNL PWR blowdown heat transfer separate-effects program, Thermal-Hydraulic Test Facility (THTF)

    1976-02-01

    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 will be obtained from the Thermal-Hydraulic Test Facility (THTF), a large nonnuclear pressurized-water loop that incorporates a 49-rod electrically heated bundle. Supporting experiments will be carried out in two additional test loops - the Forced Convection Test Facility (FCTF), a small high-pressure facility in which single heater rods can be tested in annular geometry; and an air-water loop which is used to evaluate two-phase flow-measuring instrumentation.

  20. A new formulation of the pseudocontinuous synthesis algorithm applied to the calculation of neutronic flux in PWR reactors

    A new formulation of the pseudocontinuous synthesis algorithm is applied to solve the static three dimensional two-group diffusion equations. The new method avoids ambiguities regarding interface conditions, which are inherent to the differential formulation, by resorting to the finite difference version of the differential equations involved. A considerable number of input/output options, possible core configurations and control rod positioning are implemented resulting in a very flexible as well as economical code to compute 3D fluxes, power density and reactivities of PWR reactors with partial inserted control rods. The performance of this new code is checked against the IAEA 3D Benchmark problem and results show that SINT3D yields comparable accuracy with much less computing time and memory required than in conventional 3D finite differerence codes. (Author)

  1. Detection of fuel rod leakage

    Nuclear reactor fuel rod leakage is determined by measurement of vibrational characteristics of a resilient, flexible means sealed within the upper end caps of the fuel elements. The flexible means, which is preferably a metallic diaphragm, is set into motion by the impact of an internal metal rod which is operated by an external magnetic field, thereby permitting an indication of the pressure inside a fuel element without disturbing the welded assembly. The metal rod is activated and the vibration measurements are made through the use of a special tool which fits near the end cap of the fuel element to be tested

  2. Detection of fuel rod leakage

    Nuclear reactor fuel rod leakage is determined by measurement of vibrational characteristics of a resilient, flexible means sealed within the upper end caps of the fuel elements. The flexible means, which is preferably a metallic diaphragm, is set into motion by the impact of an internal metal rod which is operated by an external magnetic field, thereby permitting an indication of the pressure inside a fuel element without disturbing the welded assembly. The metal rod is activated and the vibration measurements are made through the use of a special tool which fits near the end cap of the fuel element to be tested. 5 claims, 8 figures

  3. Simulation of the REP fuel rods thermal behaviour during transient conditions

    To control fuel rods of PWR, a laboratory of the Cea, the LEMO (Laboratory of Study and Simulation) develops a finite element code TOUTATIS. Its allows the user to calculate the deformations connected to the pellet-clad systems and hence the Pellet-Cladding Interactions (PCI) induced by unilateral contact. This paper presents the code validation from a thermal point of view. The computerized simulation is compared to experimental results CALIF and EXTRAFORT, realized to the Cea in the experimental reactor OSIRIS. (A.L.B.)

  4. Parametric study of a reactivity accident in a pressurized water reactor: control rod cluster ejection

    This research thesis concerns a class 4 accident in a PWR: the ejection of a control rod cluster from the reactor core. It aims at defining, for such an accident, the envelope values which relate the reactivity to the hot spot factor within the frame of a mode A control. The report describes the physical phenomena and their modelling during the considered transient. It presents a simple mathematical solution of the accident which shows that the main neutron parameters are the released reactivity, the delayed neutron fraction, the Doppler coefficient, and the hot spot factor. It reports a temperature sensitivity study, and discusses three-dimensional calculations of irradiation distributions

  5. Modelling activity transport behavior in PWR plant

    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)

  6. Nondestructive examination requirements for PWR vessel internals

    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

  7. Dismantling and decommissioning experience of commercial PWR

    Regarding the relatively youthness of FRAMATOME PWR's in operation none of these reactor needs to be decommissioned before 1992. However feasibility studies have been carried out by FRAMATOME for an on site entombment of active components and heavy equipments. In the past, partial dismantling of the reactor internals of the CHOOZ reactor: PWR of 320 MWe and a complete removal of the thermal shield protecting the reactor vessel were conducted successfully. After repair, the reactor power output has been upgraded of 10% and the reactor operates satisfactorily since 1970. More recently the discovery of scarce defects affecting centering pins of control guide tube located in the upper reactor internals of 900 MWe plants has initiated the construction of several ''Hot stand equipments'' for the systematic replacement of these centering pins. FRAMATOME is presently actively studying possible options consisting either to extend the plant life beyond its initial licence life, or to convert classical PWR into an advanced reactor more economical in terms of uranium consumption

  8. Pu-breeding feasibility in PWR

    This study addresses the issue of alternative pathways for breeding plutonium in a 900 MWe three loop thermal pressurized water reactor (PWR), either fueled with uranium fuel (3.5% U-235) or with mixed fuel (20% MOX). During the operation of a nuclear reactor the in-core neutron flux and the ex-core neutron flux are monitored with flux detectors. At the places where those detectors operate, the guide thimbles and the vessel wall, respectively, the neutron flux can be used to irradiate material samples. This paper investigates whether it would be possible to produce plutonium by breeding it at the walls of a PWR vessel and/or in the guide thimbles. The neutron flux in the reactor and the corresponding multi-group spectra are estimated with Monte Carlo simulations for different positions at the vessel wall of a PWR operating with either UO2 or MOX. Then the irradiation of fresh uranium samples at the vessel wall and in the guide thimbles are calculated and the isotopic composition of the irradiated samples are determined. The minimum irradiation period and the necessary minimum amount of fresh uranium to breed different grades of plutonium are derived

  9. Control rod drives

    Purpose: To enable fine positioning by using an induction motor of a simple structure as a driving source and thereby improve the reliability of control rod drives. Constitution: A step actuator is directly coupled with an induction motor, in which the induction motor is connected by way of a pulse driving control circuit to an AC power source, while the step actuator is connected to a DC power source. When a thyristor is turned ON, the motor outputs a positive torque and rotates and starts to rotate in the forward direction. When the other thyristor is turned ON, the motor is applied with braking by a reverse excitation in a manner equivalent to the change for the exciting phase sequence. When the speed is lowered to a predetermined value, braking is actuated by the torque of the step actuator and the motor stops at a zero position or balanced position. In this way, braking is actuated from the decelerating step to the stopping with no abrasion and a highly accurate positioning is possible due to the characteristics of the step actuator. (Horiuchi, T.)

  10. Nuclear fuel rod

    Purpose: To enable a wider range of output fluctuation by reducing the stress in the way of the connection between the lower end plug and the cladding tubes and thus increase the stress corrosion life. Constitution: Plurality of uranium dioxide pellets are filled in the zirconium alloy cladding tubes and the upper and lower ends are closed by zirconium alloy plugs to form nuclear fuel rods. The lower plug is provided with a hole from the inner side and in the axial direction of the plug. A structure of thermally conductive material, the conductivity of which is higher than that of the zirconium used for forming the plug, is provided in such a way that it has some clearance with the side of the said hole. By providing a hole on the lower plug and by installing a highly thermally conductive structure in it, the average temperature differential between the lower plug and the cladding tube is reduced thus reducing the thermal stress on the lower plug. (Yoshihara, Y.)

  11. Control rod drive

    Object: To provide a simple and compact construction of an apparatus for driving a drive shaft inside with a magnetic force from the outside of the primary system water side. Structure: The weight of a plunger provided with an attraction plate is supported by a plunger lift spring means so as to provide a buffer action at the time of momentary movement while also permitting the load on lift coil to be constituted solely by the load on the drive shaft. In addition, by arranging the attraction plate and lift coil so that they face each other with a small gap there-between, it is made possible to reduce the size and permit efficient utilization of the attracting force. Because of the small size, cooling can be simply carried out. Further, since there is no mechanical penetration portion, there is no possibility of leakage of the primary system water. Furthermore, concentration of load on a latch pin is prevented by arranging so that with a structure the load of the control rod to be directly beared through the scrum latch. (Kamimura, M.)

  12. Rod overpressure/lift-off testing at halden. In-pile data and analysis

    The cladding lift-off experiments at Halden aim at yielding direct data for the maximum pressure above system pressure to which a rod can be operated without causing a lasting fuel temperature increase. In addition, the tests produce a wealth of other data related to steady-state and transient thermal behaviour, PCMI, solid fission product swelling, and axial gas flow. The experiments are carried out in IFA-610 (Instrumented Fuel Assembly). The test rods contain UO2 or MOX fuel pre-irradiated in commercial LWRs to high exposures (4-5 cycles). The fuel segments are re-fabricated in the Kjeller hot cells and equipped with a fuel thermocouple and a cladding extensometer. Gas lines are attached to the end plugs and connected to a high pressure system for pressurisation of the rod with argon and a low pressure system for hydraulic diameter measurements to study cladding outward deformation and axial gas communication within the fuel rod. The test rig is equipped with vanadium and cobalt neutron detectors for monitoring the axial flux distribution and rod power. The rod instrumentation is connected to a fast scanning system for noise measurements, which are performed at regular intervals during overpressure operation. The in-pile pressure flask is surrounded by 12 booster rods with an active length of 80 cm to provide a representative fast neutron flux in the test section. The test rig is connected to an outer loop operating under PWR conditions (155 bar, 310degC). The paper summarises the test conditions and the major results of the first experiment of the test series which utilised a UO2 fuel segment pre-irradiated for four cycles in a commercial nuclear power plant to a burn-up of 52 MWd/kg UO2. The test was operated for two reactor cycles (∼4500 power hours) under representative PWR conditions. The rod was pressurised with argon, starting at 200 bar and increasing to 450 bar in five successive steps of 50 bar, while recording changes in fuel centreline temperature

  13. Zirconia thickness measurements for irradiated fuel rods: an approach to better understanding measurement error

    Non-destructive examinations (NDE) on irradiated PWR fuel rods have been performed since 1992 at the CEA/Cadarache Research Centre. Among the different controls performed, measurement of the zirconia thickness provides useful information on the axial and angular distribution of corrosion down the rod. This is necessary to compare the sensitivity of different cladding types with the creation of zirconia, as well as to detect and measure effects such as local corrosion. A dedicated apparatus based on eddy currents was used to measure the zirconia thicknesses. To verify the accuracy of our measurements, we compared the measurement results with the metallographic measurements of 39 samples. It was observed that the non-destructive measurements always overestimated the thickness of zirconia. The mean value of this systematic error was about 4 μm. We therefore tried to identify the origin of this error. We first observed that the sensor position was crucial. It must be in the exact same position for both the standard (tube section) and the rods. A poorly-positioned sensor on the rod produces overestimated measurement values. Other sources of uncertainty may also explain the difference with the exact values: first, the cladding of the standard was not irradiated. We know that some physical characteristics of cladding change during irradiation, in particular electrical conductivity. We do not know how this affects our measurement. Secondly, the rods still contained some decay heat. Thus, the temperature of the rod cladding could differ from the temperature of the standard. The electrical conductivity of the cladding and thus the eddy current response could be different. The sensor itself could also be affected by the temperature. We have performed several experiments on both heated cladding (not irradiated) and irradiated PWR fuel rods inside the hot cell. Based on the results of these tests and in agreement with our feedback, it was found that the device used in the

  14. Zirconia thickness measurements for irradiated fuel rods: an approach to better understanding measurement error

    Lacroix, B.; Martella, T.; Pras, M.; Masson-Fauchier, M. [CEA/DEN/CAD/DEC/SA3C/Legend (France); Fayette, L. [CEA/DEN/CAD/DEC/SA3C/LEMCI (France)

    2011-07-01

    Non-destructive examinations (NDE) on irradiated PWR fuel rods have been performed since 1992 at the CEA/Cadarache Research Centre. Among the different controls performed, measurement of the zirconia thickness provides useful information on the axial and angular distribution of corrosion down the rod. This is necessary to compare the sensitivity of different cladding types with the creation of zirconia, as well as to detect and measure effects such as local corrosion. A dedicated apparatus based on eddy currents was used to measure the zirconia thicknesses. To verify the accuracy of our measurements, we compared the measurement results with the metallographic measurements of 39 samples. It was observed that the non-destructive measurements always overestimated the thickness of zirconia. The mean value of this systematic error was about 4 {mu}m. We therefore tried to identify the origin of this error. We first observed that the sensor position was crucial. It must be in the exact same position for both the standard (tube section) and the rods. A poorly-positioned sensor on the rod produces overestimated measurement values. Other sources of uncertainty may also explain the difference with the exact values: first, the cladding of the standard was not irradiated. We know that some physical characteristics of cladding change during irradiation, in particular electrical conductivity. We do not know how this affects our measurement. Secondly, the rods still contained some decay heat. Thus, the temperature of the rod cladding could differ from the temperature of the standard. The electrical conductivity of the cladding and thus the eddy current response could be different. The sensor itself could also be affected by the temperature. We have performed several experiments on both heated cladding (not irradiated) and irradiated PWR fuel rods inside the hot cell. Based on the results of these tests and in agreement with our feedback, it was found that the device used in the

  15. Spacer grid for a PWR fuel assembly

    The spacer grid defines a block of square section cells each accommodating one fuel rod and is made up of interlocking flat strips welded together and made of zirconium alloy. A spring of nickel alloy is secured between each peripheral strip. The strip defining the wall of each of those cells opposite strip carries rigid bosses pressed out of the strip. The rods in those cells are gripped between bosses and spring sections. 7 figs

  16. A power control system for the rod drive coil of control element drive mechanism in pressurized water reactor

    In this paper, we propose a new type of power control system for the rod drive coil of the CEDM of the PWR NPP in order to supply more reliable DC power. The electrical modelling of the controlled rod drive coil was done by referring related documentations. The design of the proposed system is based on this electrical model satisfying the existing specification. A high power DC-DC converter scheme is adopted utilizing the SMPS technique in the design of the proposed system. In order to show the effectiveness of the proposed system, an experimental system with the capability of 3.2 K Watt was set up for a rod with four cores and some computer simulations and experimentations were carried out. The result shows a very similar tracking performance with that of the existing system to the driving command. As a result of this, the proposed method can be applied to the power control system for the rod drive coil of the CEDM of the PWR NPP. (Author). 8 refs., 1 tab., 10 figs

  17. Neutronic performance of uranium nitride composite fuels in a PWR

    Highlights: • Survey and sensitivity assembly level studies for uranium nitride composite fuels. • Composites harden the neutron spectrum and decrease the worth of control rods. • Moderator temperature coefficient is more negative, soluble boron coefficient is less negative. • Similar equilibrium core power peaking and reactivity coefficient when compared to UO2. • Illustrates “do no harm” in evaluation of candidate accident tolerant fuels. - Abstract: Uranium mononitride (UN) based composite nuclear fuels may have potential benefits in light water reactor applications, including enhanced thermal conductivity and increased fuel density. However, uranium nitride reacts chemically when in contact with water, especially at high temperatures. To overcome this challenge, several advanced composite fuels have been proposed with uranium nitride as a primary phase. The primary nitride phase is “shielded” from water by a secondary phase, which would allow the potential benefits of nitride fuels to be realized. This work is an operational assessment of four different candidate composite materials. We considered uranium dioxide (UO2) and UN base cases and compared them with the candidate composite UN-based fuels. The comparison was performed for nominal conditions in a reference PWR with Zr-based cladding. We assessed the impact of UN porosity on the operational performance, because this is a key sensitivity parameter. As composite fuels, we studied UN/U3Si5, UN/U3Si2, UN/UB4, and UN/ZrO2. In the case of UB4, the boron content is 100% enriched in 11B. The proposed zirconium dioxide (ZrO2) phase is cubic and yttria-stabilized. In all cases UN is the primary phase, with small fractions of U3Si5, U3Si5, UB4, or ZrO2 as a secondary phase. In this analysis we showed that two baseline nitride cases at different fractions of theoretical density (0.8 and 0.95) generally bound the neutronic performance of the candidate composite fuels. Performance was comparable with

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

    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)

  19. International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR

    An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental comparison basis of the severe fuel damage experiment CORA-W2. The out-of-pile experiment CORA-W2 was executed in February 1993 at he Forschungszentrum Karlsruhe. The objective of this experiment was the investigation of the behavior of a Russian type PWR fuel element (VVER-1000) during early core degradation. The main difference between a Western type and a Russian type PWR bundle is the B4C absorber rod instead of AgInCd. Measured quantities ar boundary conditions, bundle temperature, hydrogen generation and the final bundle configurations after cooldown. The ISP was conducted as a blind exercise. Boundary conditions were estimated using ATHLET-CD. Six different severe accident codes were used. The comparisons between experimental and analytical results were grouped by codes and examined separately. The thermal behavior up to significant oxidation has been predicted quite well. Larger deviations have been observed for the oxidation-induced temperature escalation, both time of onset and maximum temperature as well. The bundle behavior is greatly influenced by chemical interactions involving B4C absorber rod material, which failed relatively early at low temperature due to eutectic interaction between B4C and SS cladding as well as the SS guide tube. Regarding the complex material interaction larger differences can be recognized between calculated and measured results because of inappropriate models for material relocation and solidification processes and the lack of models describing the interactions of absorber rod materials with the fuel rods. For the total amount of H2 generated, acceptable agreement could be achieved, if the total of oxidized zirconium was calculated correctly. The oxidation of stainless steel components and B4C were not treated. In general the confidence in code predictions decreases with processing core damage. (N.T.)

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

    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. Supplementary, it is recommended to consider amendments to existing water chemistry

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

    Radioactive nuclei, such as 68Co and 60Co, 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), 60Co 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)

  2. LOFT nuclear fuel rod behavior

    An overview of the calculational models used to predict fuel rod response for Loss-of-Fluid Test (LOFT) data from the first LOFT nuclear test is presented and discussed and a comparison of predictions with experimental data is made

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

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

    2015-07-01

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

  4. Parametric study on parallel flow induced damping of PWR fuel assembly

    This paper reports on a mechanism of parallel flow-induced changes in vibrational characteristics of PWR fuel assemblies that has been studied through a series of hydraulic tests using reduced-and full-scale prototype mockups. Measured data and analytical evaluations showed the phenomenon stands on essentially the same basis as the dynamics and stability of flexible cylinders subjected to a parallel flow. In the mathematical model, the effects of rod bundle geometries and boundaries formed by walls or adjacent bundles can be exactly incorporated in the form of added mass coefficients, velocity coupling coefficients and other fluid forces. From a full scale test, it has been shown that coolant temperature has little effect up to reactor operating conditions. The updated FEM model has been verified to be applicable in describing the vibrational characteristics of from an isolated cylinder to a full scale fuel assembly in terms of the consistent properties

  5. Irradiation test for verification of PWR 48 GWd/t high burnup fuel

    Nuclear Power Engineering Corporation (NUPEC) has conducted the irradiation test for verification of the high burnup fuel performance under the sponsorship of the Ministry of Economy, Trade and Industry. (NUPEC-HB Project) As for PWR, the fuel burnup is extended by two steps. The Step I fuel (maximum fuel assembly discharge burnup: 48 GWd/t), has been utilized since 1989. And now, the preparation for the regular utilization of Step II fuel (maximum fuel assembly discharge burnup: 55 GWd/t), is being conducted. The results of pre- and post-irradiation tests on the Step I fuel irradiated in the Takahama-3 of Kansai Electric Power Co., Inc., were analyzed and evaluated. The irradiation performance of fuel rod, pellet, cladding and fuel assembly showed no remarkable difference compared with that of other published paper. Consequently the reliability and integrity of the Step I fuel was verified. (author)

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

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

  7. Nuclear fuel rod supporting arrangement

    A grid structure for holding a number of nuclear fuel rods is described. The grid structure is of the type having walls including rigidly interconnected generally rectangular metal strips, forming passageways and adapted to support nuclear fuel rods within some of the passageways. The improvement provides elongated slots intermediate and normal to the longitudinal edges of each of the strips at each intersection of the strips. The slots form openings in each corner of each passageway

  8. Distributions of Fission Products on PCI In Spent PWR Fuel Using EPMA

    The fuel specimen with 62,000 MWd/tU and the spent failed fuel rod with 53,000 MWd/tU by commercials PWR fuel were examined to compare with oxygen rich and average region at fuel-clad gap. To observe chemical behaviors and distributions of fission products on fuel-clad gap region by EPMA (Electron probe Micro-Analyzer). The results of this study can be use also in the interim storage facilities for spent fuels which were used in the Korea nuclear power plant. In addition, for comparisons of each plant’s spent fuel characteristics this data will be use as a basic material. EPMA technique offers the possibility of identifying and analyzing such phases and segregations in spent PWR fuel, although the small amount expected to be present and the background radiation, present a significant analytical challenge. The detailed characterization of spent fuel fuel-clag gap region of fission products before and after its expose from neutron is an important part of longterm storage of spent fuel. This report presents the results of EPMA examination of a spent fuel specimen with 62,000 MWd/tU performed with the aim of EPMA technique to analyses of fission products on fuel-clad gap region. (author)

  9. Stress corrosion cracking in the vessel closure head penetrations of French PWR's

    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

  10. Validation of the new code package APOLLO2.8 for accurate PWR neutronics calculations

    This paper summarizes the Qualification work performed to demonstrate the accuracy of the new APOLLO2.S/SHEM-MOC package based on JEFF3.1.1 nuclear data file for the prediction of PWR neutronics parameters. This experimental validation is based on PWR mock-up critical experiments performed in the EOLE/MINERVE zero-power reactors and on P.I. Es on spent fuel assemblies from the French PWRs. The Calculation-Experiment comparison for the main design parameters is presented: reactivity of UOX and MOX lattices, depletion calculation and fuel inventory, reactivity loss with burnup, pin-by-pin power maps, Doppler coefficient, Moderator Temperature Coefficient, Void coefficient, UO2-Gd2O3 poisoning worth, Efficiency of Ag-In-Cd and B4C control rods, Reflector Saving for both standard 2-cm baffle and GEN3 advanced thick SS reflector. From this qualification process, calculation biases and associated uncertainties are derived. This code package APOLLO2.8 is already implemented in the ARCADIA new AREVA calculation chain for core physics and is currently under implementation in the future neutronics package of the French utility Electricite de France. (authors)

  11. Status on development and verification of reactivity initiated accident analysis code for PWR (NODAL3)

    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. Comparative calculations on selected two-phase flow phenomena using major PWR system codes

    In 1988 a comparative study on important features and models in six major best estimate thermal hydraulic codes for PWR systems was implemented (Comparison of thermal hydraulic safety codes for PWR Graham, Trotman, London, EUR 11522). It was a limitation of that study that the source codes themselves were not available but the comparison had to be based on the available documentation. In the present study, the source codes were available and the capability of four system codes to predict complex two-phase flow phenomena has been assessed. Two areas of investigation were selected: (a) pressurized spray phenomena; (b) boil-up phenomena in rod bundles. As regards the first area, experimental data obtained in 1972 on the Neptunus Facility (Delft University of Technology) were compared with the results of the calculations using Athlet, Cathare, Relap 5 and TRAC-PT1 and, concerning the second area, the results of two experimental facilities obtained in 1980 and 1985 on Thetis (UKEA) and Pericles (CEA-Grenoble) were considered

  13. Simulation of small break loss of coolant accident in pressurized water reactor (PWR)

    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)

  14. Experimental investigation of reflux condensation heat transfer in PWR steam generator tubes in the presence of noncondensible gases

    Under certain circumstances in a Pressurized Water Reactor (PWR), the coolant system may be in a partially drained state and reflux condensation in the steam generator U-tubes can be the major heat removal mechanism. Noncondensable gases may be present and would degrade the heat transfer rate. If heat removal rates are insufficient, this situation could lead to core boil-off, fuel rod heatup, and eventually core damage. The Institute of Nuclear Safety System, Inc. (INSS) and the Nuclear Heat Transfer Systems Laboratory at Purdue University have begun a cooperative research program to investigate the effectiveness of reflux condensation in PWR steam generator U-tubes in the presence of noncondensable gases. The final objectives are to provide local heat transfer data for development of methods to analyze reflux condensation in PWR steam generator U-tubes and to investigate the potential for flooding. Key features of the experimental data reported herein are that they are local data under laminar steam/gas mixture and condensate film flow and they are taken from a test section with dimensions similar to an actual steam generator tube. Steady state data were obtained under various steam and air inlet flow rates and pressures. The data show the significant degrading effect of noncondensable gas on heat transfer coefficients. From the data, correlations for the reflux condensation local heat transfer coefficient and the local Nusselt number under laminar conditions were derived. These experiments are providing essential and unique fundamental data for development of methods to analyze reflux condensation

  15. Experimental investigation of reflux condensation heat transfer in PWR steam generator tubes in the presence of noncondensible gases

    Vierow, Karen; Wu, Tiejun [Purdue Univ., West Lafayette (United States); Nagae, Takashi [Institute of Nuclear Safety System, Tokyo (Japan)

    2003-07-01

    Under certain circumstances in a Pressurized Water Reactor (PWR), the coolant system may be in a partially drained state and reflux condensation in the steam generator U-tubes can be the major heat removal mechanism. Noncondensable gases may be present and would degrade the heat transfer rate. If heat removal rates are insufficient, this situation could lead to core boil-off, fuel rod heatup, and eventually core damage. The Institute of Nuclear Safety System, Inc. (INSS) and the Nuclear Heat Transfer Systems Laboratory at Purdue University have begun a cooperative research program to investigate the effectiveness of reflux condensation in PWR steam generator U-tubes in the presence of noncondensable gases. The final objectives are to provide local heat transfer data for development of methods to analyze reflux condensation in PWR steam generator U-tubes and to investigate the potential for flooding. Key features of the experimental data reported herein are that they are local data under laminar steam/gas mixture and condensate film flow and they are taken from a test section with dimensions similar to an actual steam generator tube. Steady state data were obtained under various steam and air inlet flow rates and pressures. The data show the significant degrading effect of noncondensable gas on heat transfer coefficients. From the data, correlations for the reflux condensation local heat transfer coefficient and the local Nusselt number under laminar conditions were derived. These experiments are providing essential and unique fundamental data for development of methods to analyze reflux condensation.

  16. CFD study of isothermal water flow in rod bundle with split-type spacer grid

    Batta, A.; Class, A. G.

    2014-06-01

    The design of rod bundles in nuclear application nowadays is assessed by CFD (computational fluid dynamics). The accuracy of CFD models need validation. Within the OECD/NEA benchmark MATiS-H (Measurement and Analysis of Turbulent Mixing in Sub-channels - Horizontal) a single-phase water flow in a 5x5 rod bundle is studied. In the benchmark, two types of spacer grids are tested, the swirl type and the split type, where the current study focuses on the split type spacer grid. Comparison of CFD results obtained at Karlsruhe Institut of Technology (KIT) with experimental results of KAERI (Korea Atomic Energy Research Institute) are presented. In the benchmark velocities components along selected lines downstream of the spacer grid are measured and compared to CFD results. The CFD code STAR CCM+ with the Realized k-ɛ model is used. Comparisons with experimental results show quantitative and qualitative agreement for the averaged values of velocity components. Comparisons of results to other benchmark partners using different modeling show that the selected mesh size and models for the analysis of the current case gives relatively accurate results. However, the used turbulent model (Realized k-ɛ does not capture the turbulent intensity correctly. Computation shows that the flow has very high mixing due to the spacer grid, which does not decay within the measurements domain (z/ DH =0-10 downstream of spacer grid). The same conclusion can be drawn from experimental data.

  17. Safety and Economics of High Power Density PWR with Novel Annular Fuel

    The internally and externally cooled annular fuel is a new type of fuel for PWRs that enables an increase in core power density by 50% within the same or better safety margins as the traditional solid fuel. Each assembly of traditional side dimensions has 160 annular fuel rods arranged in a 13x13 array. Even at the much higher power density, the fuel exhibits substantially lower temperatures and a MDNBR margin comparable to that of the traditional solid fuel at nominal (100%) power. Safety analyses indicate that the new annular fuel can accommodate 50% power up-rate in a PWR and still maintain adequate safety margins for a variety of transients and accidents including Loss of Flow Accident, Main Steam Line Break, Large Break Loss of Coolant Accident and Rod Ejection Accident. An economic study of 50% up-rate of an existing 1200 MW(e) PWR using the annular fuel shows that: (1) an Internal Rate of Return (IRR) on the order of 20% or more can be expected from such projects, even when accounting for uncertainties in the fuel price, electricity price inflation and cost of equipment; (2) Gradual replacement of the solid core by annular batches prior to up-rating can improve the IRR by 2.3% to 3.5% as it allows to full use of the energy in two already paid for batches of solid fuel rather than discarding them. Mixing of annular and solid fuel assemblies in one core appears feasible due to similar pressure drop characteristics of both assemblies. (authors)

  18. Qualification test of the EPR control rod drive mechanism in the full scale component test facility KOPRA

    The control rod drive mechanism (CRDM) and the mobile set consisting of rod cluster control assembly (RCC-A) of the evolutionary power reactor (EPR) had to pass a full scale qualification test in representative site conditions. The KOPRA core test section in Erlangen is precisely designed for full scale tests on nuclear core components in respect to coolant temperature and volume flow of PWR site conditions. In the test channel the complete geometry of the central core position of the reactor pressure vessel is simulated with 1:1 scale. The performance test program has led to an optimized test sequence through small adjustments in operating parameters of CRDM. The endurance test program has demonstrated that all tested components, i.e. the CRDN, the control rod driveline and the components of the drop channel are able to function properly and to meet the specification goals.

  19. A finite element method with contact for tensile analysis in fuel rods

    Elements for mechanical analysis of fuel rod of a PWR type reactor, are presented. The rod, consists basically in a cylindrical coating of zircalloy which contains pilling of UO2 pellets, is submitted to strong internal and external pressures, intense temperature gradients and neutron flux. These conditions lead several phenomena in the pellet (swelling, fracture, densification, creep) and in the cladding (embrittlement, corrosion, creep) which undergo deformations leading them to contact the restriction for the interpenetration is included in the problem without restriction by Lagrange multipliers. Considering a non-linear problem, due to the surface of contact to be not known a priori, the numerical solutions were obtained using the finite element method. (M.C.K.)

  20. Proposal of a guide to performance assessment of fuel rods for nuclear power plants

    The purpose of this paper is to present a proposal for a procedure to be adopted by the Brazilian Nuclear Energy Commission (CNEN) to evaluate the safety of fuel rods being used in nuclear power reactors in operation in the country. It should also guide the licensing process of new fuel rods designs clearly delimiting safety criteria related to its thermo-mechanical behavior. This activity is under technical collaboration of multidisciplinary design INSC BR3.01/09-BR/RA/01 Project signed between Brazil and the European Union (EU). This paper presents a first step towards establishing a CNEN standard on specific safety requirements to be met by designs of fuel elements of NPP reactors (PWR) that are operating in Brazil. (author)

  1. The pellet-cladding contact in a fuel rod and its simulation by finite elements

    A model to analyse the mechanical behavior of a fuel rod of a PWR is presented. We drew our attention to the phenomenon of pellet-pellet and pellet-cladding contact by taking advantage of a model which assumes the hypotheses of axisymmetry, elastic behavior with infinitesimal deformations and changes of the material properties with temperature. It also includes the effects of swelling and initial relocation. The problem of contact gives rise to a variational formulation which employs Lagrangian multipliers. With this approach an iterative scheme is constructed to obtain the solution. The finite element method is applied to space discretization. The model sensibility to some parameters and its performance concerning fuel rod behavior is discussed by means of numerical simulations. (author)

  2. Supplemental description of ROSA-IV/LSTF with No.1 simulated fuel-rod assembly

    Forty-two integral simulation tests of PWR small break LOCA (loss-of-coolant accident) and transient were conducted at the ROSA-IV Large-Scale Test Facility (LSTF) with the No.1 simulated fuel-rod assembly between March 1985 and August 1988. Described in the report are supplemental information on modifications of the system hardware and measuring systems, results of system characteristics tests including the initial fluid mass inventory and heat loss distribution for the primary system, and thermal properties for the heater rod materials. These are necessary to establish the correct boundary conditions of each LSTF experiment with the No.1 core assembly in addition to the system data given in the system description report (JAERI-M 84-237). (author)

  3. Impact of improved neutronic methodology on the cladding response during a PWR reactivity initiated accident

    Hursin, Mathieu, E-mail: mathieu.hursin@psi.ch [Department of Nuclear Engineering of University of California at Berkeley, 4101 Etcheverry Hall, Berkeley, CA (United States); Downar, Thomas J., E-mail: downar@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Boulevard, Ann Arbor, MI (United States); Montgomery, Robert, E-mail: Robert.Montgomery@pnnl.gov [Anatech Corp, San Diego, CA (United States)

    2013-09-15

    Highlights: • DeCART provides radial, azimuth, and axial power distribution during RIA analysis. • Coupling to FALCON is developed to evaluate the impact of such information. • Burnup calculation in the “two step” approach causes cladding load discrepancies. • The effect of azimuthal power variation has a 10% impact on the cladding load. -- Abstract: When applied to reactivity initiated accidents (RIAs) analysis, codes such as DeCART can provide a detailed radial, azimuth, and axial power distribution within a fuel rod. The work reported here is aimed at quantifying the sensitivity of the cladding thermo-mechanical response, calculated by the fuel performance code FALCON to the more accurate and detailed neutronic solution provided by DeCART for full PWR core RIA analysis. As a basis of comparison, the neutronics analysis is also performed with the U.S. NRC PARCS code, which is representative of the methodology used by the industry. Based on the DeCART solutions, several fuel rods are chosen for analysis with FALCON according to several relevant criteria. For each of the selected fuel rods, a FALCON study is performed using the boundary conditions provided by the neutronic solvers to predict the cladding response in terms of Strain Energy Density (SED) to the power pulse during the transient. The results of the analysis led to the following conclusions: • The largest impact on the cladding response can be attributed to the differences in the kinetic parameters in PARCS and DeCART. • The modeling of fuel pin exposure in the current industry standard “two step” methodology can result in some significant discrepancies in terms of SED during RIA analysis. • The effect of azimuthal power variation within a given fuel rod has a 10% impact on the SED and should be taken into consideration during RIA analysis, especially for high exposure fuel.

  4. Impact of improved neutronic methodology on the cladding response during a PWR reactivity initiated accident

    Highlights: • DeCART provides radial, azimuth, and axial power distribution during RIA analysis. • Coupling to FALCON is developed to evaluate the impact of such information. • Burnup calculation in the “two step” approach causes cladding load discrepancies. • The effect of azimuthal power variation has a 10% impact on the cladding load. -- Abstract: When applied to reactivity initiated accidents (RIAs) analysis, codes such as DeCART can provide a detailed radial, azimuth, and axial power distribution within a fuel rod. The work reported here is aimed at quantifying the sensitivity of the cladding thermo-mechanical response, calculated by the fuel performance code FALCON to the more accurate and detailed neutronic solution provided by DeCART for full PWR core RIA analysis. As a basis of comparison, the neutronics analysis is also performed with the U.S. NRC PARCS code, which is representative of the methodology used by the industry. Based on the DeCART solutions, several fuel rods are chosen for analysis with FALCON according to several relevant criteria. For each of the selected fuel rods, a FALCON study is performed using the boundary conditions provided by the neutronic solvers to predict the cladding response in terms of Strain Energy Density (SED) to the power pulse during the transient. The results of the analysis led to the following conclusions: • The largest impact on the cladding response can be attributed to the differences in the kinetic parameters in PARCS and DeCART. • The modeling of fuel pin exposure in the current industry standard “two step” methodology can result in some significant discrepancies in terms of SED during RIA analysis. • The effect of azimuthal power variation within a given fuel rod has a 10% impact on the SED and should be taken into consideration during RIA analysis, especially for high exposure fuel

  5. Two phase flow in a PWR vessel downcomer during a refill phase of LOCA

    The refill stage of the Hypothetical Loss of Coolant Accident (LOCA), of a Pressurized Water Reactor (PWR), has been the subject of considerable analysis and experimental study even through it is a very unlikely event. The Large Break Loss of Coolant Accident has been studied extensively in PWR systems to assess the effectiveness of the emergency core cooling system to maintain the fuel rods within safe temperature level. A particular phase of the transient, known as the Refill phase may be reached when the emergency core coolant, inject via the cold legs could be prevented from completely entering the core due to opposing flow of steam in the annular downcomer of the PWR vessel. Under certain conditions the upward flowing steam can hold-up the liquid and by-pass it around the downcomer annulus to the break. Experiments were performed at close to atmospheric pressure in a polycarbonate 1/10th - scale model of a typical four loop Westinghouse pressurized Water Reactor. The objective of the study was to identify hydraulic conditions and flow regimes that exist during the refill stage of a LOCA in the PWR downcomer, and to study the effects of various liquid inlet conditions (non-uniform liquid injection) into the downcomer on the flooding characteristics. The tests were performed using air and water at the working fluids to simulate the refill process under thermal equilibrium conditions. The effect of hot leg penetrations (blockages) on the flooding characteristics was also investigated and studied. The results were plotted using the Wallis J* parameter, and the mean annular circumference w (was used as the characteristic dimension). The flow regimes types and their spatial distribution in the PWR vessel downcomer for air-water counter-current flow were mapped, presented and discussed. The controlling mechanisms for the flooding were postulated and discussed. The flow pattern maps associated with various liquid modes were constructed. In addition the corresponding

  6. Isotopic Details of the Spent Catawba-1 MOX Fuel Rods at ORNL

    Ellis, Ronald James [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-01

    The United States Department of Energy funded Shaw/AREVA MOX Services LLC to fabricate four MOX Lead Test Assemblies (LTA) from weapons-grade plutonium. A total of four MOX LTAs (including MX03) were irradiated in the Catawba Nuclear Station (Unit 1) Catawba-1 PWR which operated at a total thermal power of 3411 MWt and had a core with 193 total fuel assemblies. The MOX LTAs were irradiated along with Duke Energy s irradiation of eight Westinghouse Next Generation Fuel (NGF) LEU LTAs (ref.1) and the remaining 181 LEU fuel assemblies. The MX03 LTA was irradiated in the Catawba-1 PWR core (refs.2,3) during cycles C-16 and C-17. C-16 began on June 5, 2005, and ended on November 11, 2006, after 499 effective full power days (EFPDs). C-17 started on December 29, 2006, (after a shutdown of 48 days) and continued for 485 EFPDs. The MX03 and three other MOX LTAs (and other fuel assemblies) were discharged at the end of C-17 on May 3, 2008. The design of the MOX LTAs was based on the (Framatome ANP, Inc.) Mark-BW/MOX1 17 17 fuel assembly design (refs. 4,5,6) for use in Westinghouse PWRs, but with MOX fuel rods with three Pu loading ranges: the nominal Pu loadings are 4.94 wt%, 3.30 wt%, and 2.40 wt%, respectively, for high, medium, and low Pu content. The Mark-BW/MOX1 (MOX LTA) fuel assembly design is the same as the Advanced Mark-BW fuel assembly design but with the LEU fuel rods replaced by MOX fuel rods (ref. 5). The fabrication of the fuel pellets and fuel rods for the MOX LTAs was performed at the Cadarache facility in France, with the fabrication of the LTAs performed at the MELOX facility, also in France.

  7. Simulation of leaking fuel rods

    The behaviour of failed fuel rods includes several complex phenomena. The cladding failure initiates the release of fission product from the fuel and in case of large defect even urania grains can be released into the coolant. In steady state conditions an equilibrium - diffusion type - release is expected. During transients the release is driven by a convective type leaching mechanism. There are very few experimental data on leaking WWER fuel rods. For this reason the activity measurements at the nuclear power plants provide very important information. The evaluation of measured data can help in the estimation of failed fuel rod characteristics and the prediction of transient release dynamics in power plant transients. The paper deals with the simulation of leaking fuel rods under steady state and transient conditions and describes the following new results: 1) A new algorithm has been developed for the simulation of leaking fuel rods under steady state conditions and the specific parameters of the model for the Paks NPP has been determined; 2) The steady state model has been applied to calculation of leaking fuel characteristics using iodine and noble gas activity measurement data; 3) A new computational method has been developed for the simulation of leaking fuel rods under transient conditions and the specific parameters for the Paks NPP has been determined; 4) The transient model has been applied to the simulation of shutdown process at the Paks NPP and for the prediction of the time and magnitude of 123I activity peak; 5) Using Paks NPP data a conservative value has been determined for the upper limit of the 123I release from failed fuel rods during transients

  8. Radioactive lightning rods waste treatment

    Full text: In this paper, we present alternative processes that could be adopted for the management of radioactive waste that arises from the replacement of lightning rods with attached Americium-241 sources. Lightning protectors, with Americium-241 sources attached to the air terminals, were manufactured in Brazil until 1989, when the regulatory authority overthrew the license for fabrication, commerce, and installation of radioactive lightning rods. It is estimated that, during the license period, about 75,000 such devices were set up in public, commercial and industrial buildings, including houses and schools. However, the policy of CNEN in regard to the replacement of the installed radioactive rods, has been to leave the decision to municipal governments under local building regulations, requiring only that the replaced rods be sent immediately to one of its research institutes to be treated as radioactive waste. As a consequence, the program of replacement proceeds in a low pace and until now only about twenty thousand rods have reached the waste treatment facilities The process of management that was adopted is based primarily on the assumption that the Am-241 sources will be disposed of as radioactive sealed sources, probably in a deep borehole repository. The process can be described broadly by the following steps: a) Receive and put the lightning rods in initial storage; b) Disassemble the rods and pull out the sources; c) Decontaminate and release the metal parts to metal recycling; d) Store the sources in intermediate storage; e) Package the sources in final disposal packages; and f) Send the sources for final disposal. Up to now, the disassembled devices gave rise to about 90,000 sources which are kept in storage while the design of the final disposal package is in progress. (author)

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

    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.)

  10. Void distribution analysis with high-speed x-ray CT scanner for design of PWR grid spacer

    A grid spacer of PWR fuel plays a dominant role to increase a thermal margin for safety operation of PWR since cooling effects of fuel rod are significantly promoted by fluid mixing induced by mixing vanes of grid spacers. Recently, CFD has become an available tool for the prediction of DNB performance of designed grid spacers by examining a fluid mixing effect. However, for complicated flow path like PWR fuel assembly, the current CFD is not applicable to the prediction of a two-phase slug flow due to the existence of complicated structure of interfaces between gas and liquid phases. As a result of that, an alternative approach is required to compensate the shortage of CFD applicability to the evaluation of DNB performance in the slug flow condition. In the present study, void distributions of air and water two-phase flows inside a 3x3 test rod bundle were experimentally investigated to examine the effects of grid spacers on the slug flow. A high-speed X-ray CT scanner was employed as a measurement tool that can provide with instantaneous void distributions in a cross section. Grid spacers of different types were used and the measured data were examined to investigate the differences of void distributions. Since it can be assumed that the DNB performance will be influenced by void distribution, our water DNB test data were referred to take into account the relationship between the void distributions and the performance of resistance to DNB. As a result, the detailed examinations of void profiles in flow subchannels provide the corresponding void distributions with the results of the water DNB tests. It was confirmed that the void distributions obtained by the high speed X-ray CT scanner could yield one of the available aspects from the view point of flow fields when we compare the DNB performance under two-phase slug flow condition between designed grid spacers. (author)

  11. Development of a COBRA-TF model for the PENN State University. Rod Bundle Heat Transfer program

    A research program entitled, 'Rod Bundle Heat Transfer (RBHT)' is currently being funded by the US Nuclear Regulatory Agency (NRC) and the Pennsylvania State University for investigating heat transfer during the reflood period in a typical nuclear power plant design for a large break loss of coolant accident. Information gathered by the RBHT facility will be used for improvement of the reflood heat transfer models in the NRC's thermal-hydraulic computer codes. A fully explicit sub-channel model of the RBHT facility has been developed using the COBRA-TF best estimate code to make pre-test calculations to aide in facility design. The model predictions help ensure all relevant heat transfer phenomena are captured during an RBHT reflood experiment. Also, the COBRA-TF model confirms previous thermal radiation calculations, which show the effectiveness of the outer row of rods in the 7 x 7 bundle in thermally shielding the inner 5 x 5 rods from the housing. Preliminary calculations are made for a flooding rate of 2.54 centimeters per second to determine the quench location at the peak cladding temperature (PCT) time. With an estimate of the quench location at PCT, the designers can better locate instrumentation to capture quench information at and before turn around time. (author)

  12. PWR Core 2 Project accident analysis

    The various operations required for receipt, handling, defueling and storage of spent Shippingport PWR Core 2 fuel assemblies have been evaluated to determine the potential accidents and their consequences. These operations will introduce approximately 16,500 kilograms of depleted natural uranium (as UO2), 139 kilograms of plutonium, 2.8 megacuries of mixed fission products, and 14 kilograms of Zircaloy-4 (cladding and hardware) into the 221-T Canyon Building. Event sequences for potential accidents that were considered included (1) leaking fuel assemblies, (2) fire and explosion, (3) loss of coolant or cooling capability, (4) dropped and/or damaged fuel assemblies, and extrinsic occurrences such as loss of services, missile impact, and natural occurrences (e.g., earthquake, tornado). Accident frequencies were determined by formal analysis to be very low. Accident consequences are greatly mitigated by the safety and containment features designed into the fuel modules and shipping cask, the long cooling time since reactor discharge, and the redundant safety features designed into the facilities, equipment, and operating procedures for the PWR Core 2 Project. Possible hazards associated with the handling of these fuels have been considered and adequate safeguards and storage constraints identified. The operations of M-160 cask unloading and module storage will not involve identifiable risks as great or significantly greater than those for comparable licensed nuclear facilities, nor will hazards or risks be significantly different from comparable past 221-T Plant programs. Therefore, it is concluded that the operations required for receipt, handling, and defueling of the M-160 cask and for the storage and surveillance of the PWR Core 2 fuel assemblies at the 221-T Canyon Building can be performed without undue risk to the safety of the involved personnel, the public, the environment or the facility

  13. PWR integral tie plate and locking mechanism

    A locking mechanism for securing an upper tie plate to the tie rods of a nuclear fuel bundle is described. The mechanism includes an upper tie plate assembly and locking sleeves fixed to the ends of the tie rods. The tie plate is part of the upper tie plate assembly and is secured to the fuel bundle by securing the entire upper tie plate assembly to the locking sleeves fixed to the tie rods. The assembly includes, in addition to the tie plate, locking nuts for engaging the locking sleeves, retaining sleeves to operably connect the locking nuts to the assembly, a spring biased reaction plate to restrain the locking nuts in the locked position and a means to facilitate the removal of the entire assembly as a unit from the fuel bundle

  14. 14C Behaviour in PWR coolant

    Although 14C 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. 14C can exist in a range of possible chemical forms: inorganic carbon compounds (probably mainly CO2), elemental carbon, and organic compounds such as hydrocarbons. This paper presents results from a preliminary survey of the possible reactions of 14C in PWR coolant. The main conclusions of the study are: - A combination of thermal and radiolytic reactions controls the chemistry of 14C in reactor coolant. A simple chemical kinetic model predicts that CH3OH would be the initial product from radiolytic reactions of 14C following its formation from 17O. CH3OH is predicted to arise as a result of reactions of OH. with CH4 and CH3, and it persists because there is no known radiation chemical reduction mechanism. - Thermodynamic considerations show that CH3OH can be thermally reduced to CH4 in PWR conditions, although formation of CO2 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 CH4 is the dominant form in PWR and CO2 in BWR. This implies that radiation chemistry may be important in determining the speciation. - Addition of acetate does not affect the amount of 14C 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 12C compared with organic material released from CVCS resins, although the importance of this may depend on whether that is predominantly soluble material or suspended

  15. Industrywide survey of PWR organics. Final report

    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

  16. Transient study of a PWR pressurizer

    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. Modelling of sucker rod string

    Hojjati, M.H. [Mazandaran Univ., (Iran, Islamic Republic of). Dept. of Mechanical Engineering; Lukasiewicz, S.A. [Calgary Univ., AB (Canada). Dept. of Mechanical and Manufacturing Engineering

    2005-12-01

    Rod pumping is used extensively in the oil well industry as a method of artificial lift. In order to analyze the performance of oil wells, the force and displacement at the polished rod are measured using a dynamometer. The data is applied to the boundary conditions when calculating the forces and displacement at the bottom of the rod string that defines the conditions of the pump, pumping effectiveness and production rate. This study proposed a transfer matrix method to model the dynamic behavior of the sucker string rod. The main reason for developing the method was to simplify the currently used mathematical method with a simple matrix operation in which the bottom-hole force-displacement values are obtained as a product of data vectors at the polished rod end by a transfer matrix. The problem was solved using D'Alembert's systems solution equation and the adaptive filter matrix method. The proposed method reduces calculation time because a more efficient matrix operation is used without losing accuracy. This study showed that it is possible to use the transfer matrix to calculate load-displacement relations a hundred or more times in one stroke, which is beneficial when developing tools to control oil wells, such as wellhead controllers. 9 refs., 3 tabs., 8 figs.

  18. The Third ATLAS ROD Workshop

    Poggioli, L.

    A new-style Workshop After two successful ATLAS ROD Workshops dedicated to the ROD hardware and held at the Geneva University in 1998 and in 2000, a new style Workshop took place at LAPP in Annecy on November 14-15, 2002. This time the Workshop was fully dedicated to the ROD-TDAQ integration and software in view of the near future integration activities of the final RODs for the detector assembly and commissioning. More precisely, the aim of this workshop was to get from the sub-detectors the parameters needed for T-DAQ, as well as status and plans from ROD builders. On the other hand, what was decided and assumed had to be stated (like EB decisions and URDs), and also support plans. The Workshop gathered about 70 participants from all ATLAS sub-detectors and the T-DAQ community. The quite dense agenda allowed nevertheless for many lively discussions, and for a dinner in the old town of Annecy. The Sessions The Workshop was organized in five main sessions: Assumptions and recommendations Sub-de...

  19. Assessment of Fuel Rod Failure Thresholds for Reactivity Initiated Accidents

    sensitivity studies indicate that the pulse width may have a significant impact on the failure enthalpy, at least for pulses narrower than 50 ms. Clad corrosion seems on the other hand to be of minor importance to the failure enthalpy, as long as the clad oxide layer is non-spalled and possible hydrides in the material are uniformly distributed. However, for cladding tubes with spalled oxide, the ductility of the material may be dramatically reduced as a consequence of non-uniform hydride precipitation, and the failure threshold significantly lower. The calculated failure enthalpy for PWR fuel rods with spalled oxide, subjected to the postulated HZP REA, is approximately 350 J/gUO2 at a fuel burnup of 65 MWd/kgU. In conclusion, the performed analyses indicate that a common fuel rod failure threshold for HZP REA and CZP CRDA, expressed in terms of allowable fuel enthalpy with respect to fuel burnup, is feasible, provided that the threshold is applied to fuel rods with non-spalled clad oxide

  20. Grid to rod fretting evaluation. Methodology of testing in the hermes loops

    Using the HERMES loops of the CEA/Nuclear Reactor Division as an experimental support, an overall methodology of fuel qualification has been developed. This qualification of new designs is been imposed by the development, by fuel assemblies' designers, of assembly prototypes in order to increase to burn up fraction. These developments induce changes in hydraulic resistance of assemblies or changes in the mechanical behavior of the structure. Besides, the diversification of fuel suppliers leads nuclear power plant operators to deal with compatibility inside heterogeneous cores. The prototype 1-scale assemblies are certified in the HERMES P loop in terms of pressure drop coefficients, scram and wear evaluation obtained after an endurance test under representative PWR conditions. After the endurance test, the fuel rods are extracted from the assembly and grid-to-rod fretting generated during the test under PWR conditions is evaluated. The second part of the qualification consists in flow induced vibrations' evaluation in the HERMES T loop in a bi-assembly configuration using laser vibrometry and laser velocimetry

  1. Refabricated and instrumented fuel rods

    Nuclear Fuel for power reactors capabilities evaluation is strongly based on the intimate knowledge of its behaviour under irradiation. This knowledge can be acquired from refabricated and instrumented fuel rods irradiated at different levels in commercial reactors. This paper presents the development and qualification of a new technique called RECTO related to a double-instrumented rod re-fabrication process developed by CEA/LECA hot laboratory facility at CADARACHE. The technique development includes manufacturing of the properly dimensioned cavity in the fuel pellet stack to house the thermocouple and the use of a newly designed pressure transducer. An analytic irradiation of such a double-instrumented fuel rod will be performed in OSIRIS test reactor starting October 2004. (Author)

  2. Advanced gray rod control assembly

    Drudy, Keith J; Carlson, William R; Conner, Michael E; Goldenfield, Mark; Hone, Michael J; Long, Jr., Carroll J; Parkinson, Jerod; Pomirleanu, Radu O

    2013-09-17

    An advanced gray rod control assembly (GRCA) for a nuclear reactor. The GRCA provides controlled insertion of gray rod assemblies into the reactor, thereby controlling the rate of power produced by the reactor and providing reactivity control at full power. Each gray rod assembly includes an elongated tubular member, a primary neutron-absorber disposed within the tubular member said neutron-absorber comprising an absorber material, preferably tungsten, having a 2200 m/s neutron absorption microscopic capture cross-section of from 10 to 30 barns. An internal support tube can be positioned between the primary absorber and the tubular member as a secondary absorber to enhance neutron absorption, absorber depletion, assembly weight, and assembly heat transfer characteristics.

  3. Prototypical Rod Consolidation Demonstration Project

    The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 2 discusses the following topics: Fuel Rod Extraction System Test Results and Analysis Reports and Clamping Table Test Results and Analysis Reports

  4. Prototypical Rod Consolidation Demonstration Project

    The objective of Phase III of the Prototypical Rod Consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod Consolidation System as described in the NUS Phase II Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase III effort the system was tested on a component, subsystem, and system level. Volume IV provides the Operating and Maintenance Manual for the Prototypical Rod Consolidation System that was installed at the Cold Test Facility. This document, Book 5 of Volume IV, discusses: Corrective maintenance procedures; Calibration procedures; Surveillance procedures; Equipment changeover procedures; Decontamination procedures; Recovery procedures; and Cable schedule

  5. Preliminary Study on the Fretting Wear Behaviors of a Duel Cooled Fuel Rod

    Based on MIT's concept, an innovative fuel development project was launched by KAERI that a substantial power up-rating could be realized by introducing an internally and externally double cooled annular fuel for current PWR reactors. In order to apply this duel cooled fuel to an OPR 1000 reactor system, geometrical features of structural parts in a fuel assembly should be changed except an overall dimension of a fuel assembly. Typical changes are summarized as fuel rod diameter and weight, shape and position of a spacer grid spring, etc. When considering a duel cooled fuel rod, its vibration characteristic and fretting behavior should be verified because the modified shape and dimension of spacer grid spring, fuel rod diameter and weight, number of spacer grid assembly are closely related to a flow-induced vibration in a duel cooled fuel assembly. In this study, based on FIV test results of 4x4 fuel assembly, fretting wear tests of an outer duel cooled fuel rod were performed by using an embossing type spacer grid spring that could adjust its spring stiffness. The discussion was focused on the evaluation of the optimized spring stiffness and spring position in 1x1 cell by analyzing the fretting wear results. (authors)

  6. Improvement of input parameters for the estimation of fuel rod temperature in dry transport cask

    A typical PWR spent fuel bundle has a 17 x 17 rod array, and an analysis requires a very long computation time and a vast amount of memory. Therefore, we applied the lumped fuel bundle analysis approach with the homogenized method to estimate the fuel cladding temperature efficiency. Thermal analysis results for lumped fuel bundles showed an excessive radiative heat transfer, and we applied an emissivity modification factor to compensate for this radiation effect. The value of the factor decreased as the number of the rods in the homogenized array decreased.. For the lumped 8 x 8 array, the best emissivity modification factor was shown to be 0.40. The rod emissivity of 0.8 is generally recommended to be used in COBRA-SFS[D. R. Rector et al.] calculations. Therefore, we can use the modified rod emissivity of 0.32 for lumped 8 x 8 array. There are good agreements between the results from lumped 8 x 8 array bundle and the results from real 17 x 17 array bundle. By homogenization, we can increase the computational speed substantially, as well as reduce the requirements on computer memory and space. (authors)

  7. The ''THERMOST'' for analysing thermo-structural behaviour of LWR fuel rod under PCI conditions

    As one of the methods for evaluating the fuel rod performances under power ramping or load following operations, the combined ''FROST'' and ''THERMOST'' system has been developed and being brought into practical use. The former had already been presented at Blackpool Meeting in 1978, and the latter is going to be presented in this paper. The major purpose of the THERMOST is to analyse very detailed thermal and structural fuel behaviours in a rather localized part of fuel rod whereas the FROST deals with whole-rod-wide general performances. The code handles 2-dimensional thermal and structural analyses simultaneously by using finite element method, in axial section wide or in lateral section wide. It consists of a fundamental FEM system of generalized constitution and its surrounding subroutine system which characterizes fuel behaviours such as temperature distribution, thermal expansion, elastoplasticity, creep, cracking, swelling, growth, etc. Thermal analysis is handled by heat conduction and heat transfer elements (6 kinds) and structural analysis by axisymmetric ring and lateral plane elements (6 kinds). Boundary problems such as contact, friction and cracking are treated by gap and crack elements. A sample calculation of PCI performance on a PWR fuel rod under ramping condition is presented with some inpile test data. (author)

  8. CECP, Decommissioning Costs for PWR and BWR

    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)

  9. Navy lifts veil on PWR research

    The author describes the experience of Rolls Royce in developing nuclear reactors for the Navy. Reference is made to the commissioning of HMS Sceptre in February 1978, Britain's 14th nuclear submarine. This event coincided with a decision to lift the veil somewhat on a Research and Development programme that has remained secret for nearly 20 years. Factors that have inhibited progress in this field are mentioned. One of these factors has been the high cost of marine nuclear propulsion systems, tending to limit interest to very large vessels or some special purpose craft. Another factor has been slowness to develop universally acceptable safety criteria, to allow for free and ready access of nuclear vessels to ports. A third factor has been the military origins of much of the development work. A new factor that has arisen recently is the development of the Westinghouse PWR (pressurised water reactor) for marine use in the UK. This has involved collaboration with the US Westinghouse Electric Corporation. Rolls Royce and Associates were chosen to manage this work, which is here described, including the first PWR to be designed and built in Britain and incorporated into a submarine (HMS Vulcan). Much of the design work has been concerned with development of the reactor core and increasing the endurance of the vessel between refuellings. Another aspect was less noise and vibration. Costs of this work are stated, and new test facilities are described. (U.K.)

  10. Workers doses in central European PWR NPPs

    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. Analysis of reciprocating compressor piston rod failures

    Tripp, H.A.; Drosjack, M.J.

    1984-02-01

    This report presents the analysis of five piston rod failures which occurred on reciprocating compressors. Calculations are shown for rod stress which includes nominal rod loading sources as well as additional loads due to unusual pressure losses in the compressor valves, flexure of the rods due to misalignment, and manufacturing errors. The additional loads were incorporated on the basis of field measurements. The stress values are used with Baquin's equation to produce fatigue life curves for the rods. Based on the calculations, recommendations for modified rods were made. The calculation procedures are described in a manner which will permit their application to other reciprocating compressors.

  12. Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code

    Trombetta, Debora M.; Su, Jian, E-mail: dtrombetta@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Chirayath, Sunil S., E-mail: sunilsc@tamu.edu [Department of Nuclear Engineering and Nuclear Security Science and Policy Institute, Texas A and M University, TX (United States)

    2015-07-01

    Computational models of reactors are increasingly used to predict nuclear reactor physics parameters responsible for reactivity changes which could lead to accidents and losses. In this work, preliminary results for criticality coefficient calculation using the Monte Carlo transport code MCNPX were presented for a small PWR. The computational modeling developed consists of the core with fuel elements, radial reflectors, and control rods inside a pressure vessel. Three different geometries were simulated, a single fuel pin, a fuel assembly and the core, with the aim to compare the criticality coefficients among themselves.The criticality coefficients calculated were: Doppler Temperature Coefficient, Coolant Temperature Coefficient, Coolant Void Coefficient, Power Coefficient, and Control Rod Worth. The coefficient values calculated by the MCNP code were compared with literature results, showing good agreement with reference data, which validate the computational model developed and allow it to be used to perform more complex studies. Criticality Coefficient values for the three simulations done had little discrepancy for almost all coefficients investigated, the only exception was the Power Coefficient. Preliminary results presented show that simple modelling as a fuel assembly can describe changes at almost all the criticality coefficients, avoiding the need of a complex core simulation. (author)

  13. Modeling and simulation of the core in PWR nuclear power plant based on PAnySimu

    Modeling and simulation analysis on the core of PWR nuclear power plant based on PAnySimu Simulation Support System. We have divided the core into five models by studying the Unit 3/4's actual core structure of Ling Ao Phase Ⅱ, which are, power calculation, calculation of core transmission, control rod reactions, reactivity feedback calculation and poison calculation. On this basis, analyses the core neutron flux, consider the influence of the control rod position, fuel and moderator temperature, xenon and samarium poisoning, boron concentration on neutron Biomass. The various modules of algorithms and primitives are defined by using PAnySimu simulation system. Then prepare the corresponding module program in the C++ Builder 5.0 environment, and storage and debug algorithm. After running the module, build core model simulation system in accordance with logical relations. At last do the dynamic simulation separately on the reaction of individual modules and systems under other circumstances of disturbance. The analysis of the real-time dates shows that results are reasonable. For some hard core within the experimental operation in the other, the simulation method to obtain experimental data, is of great significance. (authors)

  14. Application of fiberglass sucker rods

    Gibbs, S.G. (Nabla Corporation (US))

    1991-05-01

    Fiberglass sucker rods are assuming a place in artificial-lift technology. This paper briefly describes the manufacturing process and gives some design and operational hints for practical applications. It also describes some mathematical modeling modifications needed for fiberglass wave-equation design programs.

  15. Application of fiberglass sucker rods

    Fiberglass sucker rods are assuming a place in artificial-lift technology. This paper briefly describes the manufacturing process and gives some design and operational hints for practical applications. It also describes some mathematical modeling modifications needed for fiberglass wave-equation design programs

  16. Turbodrill rod angular velocity indicator

    Rogachev, O.K.; Belozerova, L.P.; Konenkov, A.K.

    1984-01-01

    This paper outlines shortcomings of existing types of telemetry systems which resulted in production of the IChT-1 unit. Unit is intended for control of angular velocity of serially produced turbodrill rods, during drilling of wells up to 5000 m deep, and bottomhole temperatures to 100C. The paper provides a detailed description and diagrams for installing this unit.

  17. Flow resistance in rod assemblies

    The general form of relation between the resistance force and the velocity vector, resistance tensor structure and possible types of anisotropy in the flow thorough such structures as rod or tube assemblies are under discussion. Some questions of experimental determination of volumetric resistance force tensor are also under consideration. (author)

  18. Basic information about development and construction of a PWR

    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.)

  19. Control rods in LMFBRs: a physics assessment

    This physics assessment is based on roughly 300 control rod worth measurements in ZPPR from 1972 to 1981. All ZPPR assemblies simulated mixed-oxide LMFBRs, representing sizes of 350, 700, and 900 MWe. Control rod worth measurements included single rods, various combinations of rods, and Ta and Eu rods. Additional measurements studied variations in B4C enrichment, rod interaction effects, variations in rod geometry, neutron streaming in sodium-filled channels, and axial worth profiles. Analyses were done with design-equivalent methods, using ENDF/B Version IV data. Some computations for the sensitivities to approximations in the methods have been included. Comparisons of these analyses with the experiments have allowed the status of control rod physics in the US to be clearly defined

  20. Control rods in LMFBRs: a physics assessment

    McFarlane, H.F.; Collins, P.J.

    1982-08-01

    This physics assessment is based on roughly 300 control rod worth measurements in ZPPR from 1972 to 1981. All ZPPR assemblies simulated mixed-oxide LMFBRs, representing sizes of 350, 700, and 900 MWe. Control rod worth measurements included single rods, various combinations of rods, and Ta and Eu rods. Additional measurements studied variations in B/sub 4/C enrichment, rod interaction effects, variations in rod geometry, neutron streaming in sodium-filled channels, and axial worth profiles. Analyses were done with design-equivalent methods, using ENDF/B Version IV data. Some computations for the sensitivities to approximations in the methods have been included. Comparisons of these analyses with the experiments have allowed the status of control rod physics in the US to be clearly defined.

  1. Digitization on control rod position indication system

    This paper introduces the design mechanism, system structure of Control Rod Position Indication System and the application of Programmable Logic Controller (PLC) on Control Rod Position Indication System. (authors)

  2. The KWU fission gas release model for LWR fuel rods: Description, data base and parametric study of the model

    The KWU model for fission gas release of LWR fuel is based on the physical processes of fission gas release as reported in the published literature and on KWU post irradiation examination results. The model is composed of two different submodels which predict the steady state and the transient fission gas release, respectively. 1) The steady state submodel can be divided into two main parts. Part 1: The fission gas produced is retained in the UO2-matrix up to a certain saturation concentration, and all fission gas exceeding this matrix concentration is collected at grain boundaries. The temperature and burnup dependent saturation concentration of the matrix has been taken from experimental results published in the literature. Part 2: The rate of fission gas release df/dt to the void volume is assumed to be proportional to the gas inventory g at the grain boundaries: df/dt=K.g. The factor K depends on temperature, burnup and open porosity. 2) The submodel for transient fission gas release is presently based on the assumption that transient release is caused by grain boundary separations due to the growth of grain surface bubbles. The transient gas release calculated in the model depends on the inventory g of fission gas retained at grain boundaries and on the power increase Δq-prime during the transient. The fission gas release model is implemented in the KWU fuel rod computer code CARO and calibrated against measured fission gas release values of approximately 100 KWU fuel rods: PWR and BWR fuel rods with burnups up to 40 MWd/kg(U), unpressurized and pre-pressurized fuel rods, rods with mixed oxide fuel and test rods with center line temperature up to 2000 deg. C, rods under normal operation and rods with a transient at the end of operation. A parametric study demonstrates the characteristic behaviour of the model. (author)

  3. Elaboration and qualification of a reference calculation routes for the absorbers in the PWR reactors

    The general field in which this work takes place is the field of the accuracy improvement of neutronic calculations, required to operate Pressurized Water Reactors (PWR) with a better precision and a lower cost. More specifically, this thesis deals with the calculation of the absorber clusters used to control these reactors. The first aim of that work was to define and validate a reference calculation route of such an absorber cluster, based on the deterministic code Apollo 2. This calculation scheme was then to be checked against experimental data. This study of the complex situation of absorber clusters required several intermediate studies, of simpler problems, such as the study of fuel rods lattices and the study of single absorber rods (B4C, AIC, Hafnium) isolated in such lattices. Each one of these different studies led to a particular reference calculation route. All these calculation routes were developed against reference continuous energy Monte-Carlo calculations, carried out with the stochastic code TRIPOLI14. They were then checked against experimental data measured during french experimental programs, undertaken within the EOLE experimental reactor, at the Nuclear Research Center of Cadarache: the MISTRAL experiments for the study of isolated absorber rods and the EPICURE experiments for the study of absorber clusters. This work led to important improvements in the calculation of isolated absorbers and absorber clusters. The reactivity worth of these clusters in particular, can now be obtained with a great accuracy: the discrepancy observed between the calculated and the experimental values is less than 2.5 %, and then slightly lower than the experimental uncertainty. (author)

  4. Evaluation of tight-pitch PWR cores

    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 - 235U/UO2 : Pu/ThO2 : 233U/ThO2 - and the conventional recycle-mode uranium system - 235U/UO2 : Pu/UO2. 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

  5. A pressure drop model for PWR grids

    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)

  6. Zebra: An advanced PWR lattice code

    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)

  7. Zebra: An advanced PWR lattice code

    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)

  8. Evaluation model for PWR irradiated fuel

    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)

  9. Crevice chemistry control in PWR steam generators

    To establish a basis for predicting and eventually controlling crevice solution chemistry in PWR steam generators, hideout tests were performed at several units. Results indicated that impurity hideout rates varied with the species and with bulk water concentration. Field evaluations of crevice impurity inventory models based on the hideout rate data indicated that further model refinements were necessary, e.g., more frequent quantification of the relation of hideout rates and bulk water concentration. An alternate crevice inventory model based on a real-time mass balance approach also began to be pursued. Modeling results currently are being used at several PWRs to establish a chloride injection rate consistent with development of a near neutral crevice solution to minimize IGA/SCC. Hideout return data are being used to independently establish predictions of crevice chemistry and to substantiate the hideout rate and mass balance model predictions

  10. The underclad cracking in PWR reactor vessels

    The article describes the kind of cracking which can occur under the stainless steel cladding during the manufacturing process of PWR vessels: - cold cracking recently found in France on vessel nozzles-reheat cracking discovered some ten years ago in particular in Germany and in USA. Methods of examination for underclad cracking are put forward, together with results obtained on vessel nozzles of units currently being built in Belgium. Some nozzles are affected by the phenomenon of reheat cracking, whilst the hypothesis of cold cracking, which had been proposed because of the similar situation found in France should probably be abandoned. On the basis of the investigations and studies made, it is established that the cracking involved does not jeopardize the integrity of the vessels during their life time. (author)

  11. The material analysis for PWR primary equipment

    The primary equipment in pressurized water reactor includes reactor pressure vessel, reactor coolant piping, steam generator, pressurizer, and reactor coolant pump casing, etc., which form the pressure boundary of the primary loop. These primary equipment are all pressure vessels of QA Class 1, Safety-related Class 1, and Aseismatic Category 1. Under high temperature, high pressure and neutron irradiation, the requirements for the base material and welding properties of these pressure vessels are very high, so as to ensure the long-term stable operation of nuclear power plant. The base material and welding properties of these pressure vessels are analyzed and discussed according to ASME B and P Code, which can be as a reference for base material selection of PWR pressure vessels. (authors)

  12. Subcooled decompression analysis in PWR LOCA

    The thermo-hydraulic behavior of the coolant in the primary system of a nuclear reactor is important in the core heat transfer analysis during a hypothetical loss-of-coolant accident (LOCA). The heat transfer correlations are strongly dependent on local thermo-hydraulic conditions of the coolant. The present work allows to calculate such thermo-hydraulic behavior of the coolant during subcooled decompression in PWR LOCA by solving the mass, momentum, and energy conservation equations by the method of characteristics. Detailed studies were made on the transient coolant outflow at the pipe rupture and the effect of frictional loss and heat addition to the coolant on the decompression. Based on the studies, a digital computer code, DEPCO-MULTI, has been prepared and numerical results are compared with the ROSA (JAERI) and the LOFT (NRTS) semiscale test data with various coolant pressures, temperatures, pipe break sizes, and complexity of flow geometry. Good agreement is generally obtained

  13. Recriticality risk in PWR spent fuel pools

    In this paper we investigated the situation in a PWR Spent Fuel Pool (SFP) following a long-term loss of power / loss of cooling accident. In the SFP there is a large amount of water with soluble boron between the fuel assemblies. There may be a problem from the point of view of criticality safety if the water of the SFP starts to boil and evaporate. A thermal-hydraulic analysis was performed using a simplified model of the SFP. The thermal-hydraulic analysis shows that in all cases a chaotic boiling phenomenon develops. This indicates that even if there is an issue of (near-)criticality, it will have a very intermittent nature. The multiplication factor of the SFP was evaluated with a Monte Carlo calculation. The neutronic analysis was performed for several representative cooling situations. In all cases, the system remains (deeply) subcritical. (author)

  14. Exponential experiments on PWR spent fuel assemblies

    An Exponential experiment system which is composed of a neutron detector, a signal analysis system and a neutron source, Cf-252 has been installed in order to experimentally determine the neutron effective multiplication factor for PWR spent fuel assembly. The axial background neutron flux is measured as a preliminary performance test. From the results, the spacer grid position is determined to be consistent with the design specifications within a 2.3% relative error. The induced fission neutron for four of the assemblies is also measured by scanning the neutron source, Cf-252 or the neutron detector. The exponential decay constants have been evaluated by the application of Poisson regression to the net induced fission neutron counts. It was revealed that the average exponential decay constants for the C15, J14, G23 and J44 assemblies were determined to be 0.130, 0.127, 0.125 and 0.121, respectively. (author)

  15. Stochastic optimization of loading pattern for PWR

    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

  16. Solitary waves on nonlinear elastic rods. II

    Sørensen, Mads Peter; Christiansen, Peter Leth; Lomdahl, P. S.;

    1987-01-01

    In continuation of an earlier study of propagation of solitary waves on nonlinear elastic rods, numerical investigations of blowup, reflection, and fission at continuous and discontinuous variation of the cross section for the rod and reflection at the end of the rod are presented. The results...

  17. Fuel followed control rod installation at AFRRI

    Fuel Followed Control Rods (FFCRs) were installed at the Armed Forces Radiobiology Research Institute's 1 MW TRIGA Reactor. The procedures for obtaining, shipping, and installing the FFCRs is described. As part of the FFCR installation, the transient rod drive was relocated. Core performance due to the addition of the fuel followed control rods is discussed. (author)

  18. Process and apparatus for controlling control rods

    This process and apparatus is characterized by 2 methods, for examination of cluster of nuclear control rods. Foucault current analyzer which examines fraction by fraction all the control rods. This examination is made by rotation of the cluster. Doubtful rods are then analysed by ultrasonic probe

  19. Leaf spring puller for nuclear fuel rods

    Fogg, J.L.

    1981-11-03

    A fuel rod puller in the form of a collet for pulling fuel rods from a storage area into grids of a nuclear reactor fuel assembly. The rod puller moves longitudinally through the grids to a storage area where projections on the end of leaf springs grasp onto an end plug in a fuel rod. Drive apparatus then pulls the rod puller and connected fuel rod from the storage area into the fuel assembly grids. The rod puller includes an outer tube having leaf springs on one end thereof in one modification, mounted within the outer tube is a movable plunger which acts to urge the leaf springs outwardly to a position to permit passing or with the end of a end plug. Upon withdrawal of the plunger, the leaf springs move into a groove formed in the end of a fuel rod end plug, and the fuel rod subsequently is pulled into the fuel assembly grids. In another modification, the leaf springs on the outer rod are biased in an outward direction and a longitudinally movable tube on the outer rod is moved in a direction to contract the leaf springs into a position where the projections thereof engage the groove formed in a fuel rod end plug.

  20. Rod and lamellar growth of eutectic

    M. Trepczyńska - Łent

    2010-01-01

    The paper presents adaptation problem of lamellar growth of eutectic. The formation of rod eutectic microstructure was investigated systematically. A new rod eutectic configuration was observed in which the rods form with elliptical cylindrical shape. A new interpretation of the eutectic growth theory was proposed.

  1. Rod and lamellar growth of eutectic

    M. Trepczyńska-Łent

    2010-04-01

    Full Text Available The paper presents adaptation problem of lamellar growth of eutectic. The formation of rod eutectic microstructure was investigated systematically. A new rod eutectic configuration was observed in which the rods form with elliptical cylindrical shape. A new interpretation of the eutectic growth theory was proposed.

  2. Detecting small flaws in fuel rods with sophisticated Eddy Current testing and single rod sipping

    AREVA has profound experience in efficient methods for finding defective irradiated fuel rods: - Eddy Current (EC) testing of single fuel rods - Sipping of fuel assemblies. In order to further improve the efficiency, AREVA developed new techniques: A. New sophisticated EC for detecting Small Defects in Irradiated Fuel Rod Claddings (SDIRC-EC) B. Single rod sipping (orig.)

  3. Radiation embrittlement of PWR vessel supports

    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

  4. Effects of cross sections tables generation and optimization on rod ejection transient analyses

    Highlights: • Different cross-section libraries are applied to a rod ejection transient benchmark. • Effects of the optimization of the library grid-point distribution are assessed. • Effects of the library generation are assessed by comparison with other solutions. • Interpolation errors contribute to neutronics uncertainties in modeling transients. - Abstract: Best estimate analysis of rod ejection transients requires 3D kinetics core simulators. If they use cross sections libraries compiled in multidimensional tables, interpolation errors – originated when the core simulator computes the cross sections from the table values – are a source of uncertainty in k-effective calculations that should be accounted for. Those errors depend on the grid covering the domain of state variables and can be easily reduced, in contrast with other sources of uncertainties such as the ones due to nuclear data, by choosing an optimized grid distribution. The present paper assesses the impact of the grid structure on a PWR rod ejection transient analysis using the coupled neutron-kinetics/thermal-hydraulics COBAYA3/COBRA-TF system. For this purpose, the OECD/NEA PWR MOX/UO2 core transient benchmark has been chosen, as material compositions and geometries are available, allowing the use of lattice codes to generate libraries with different grid structures. Since a complete nodal cross-section library is also provided as part of the benchmark specifications, the effects of the library generation on transient behavior are also analyzed. Results showed large discrepancies when using the benchmark library and own-generated libraries when compared with benchmark participants’ solutions. The origin of the discrepancies was found to lie in the nodal cross sections provided in the benchmark

  5. Measurement of pressure drops in prototypic BWR and PWR fuel assemblies in the laminar regime - Pressure drop measurement of laminar air flow in prototypic BWR and PWR fuel assemblies

    Laminar pressure drops in nuclear fuel assemblies are of interest for evaluating complete loss-of-coolant accident scenarios in spent fuel pools and for performance analyses of dry storage casks. To the knowledge of the authors, this study represents the first attempt to directly quantify pressure losses in prototypic fuel assemblies in the laminar regime. Two commercial fuel assemblies were examined including a 17x17 PWR and a 9x9 BWR. The assemblies were tested in the laminar regime with Reynolds numbers ranging from 10 to 1000, based on the average assembly velocity and hydraulic diameter. Pressure drop measurements were made across individual bundle spans and grid spacers in the mock fuel assemblies using high-sensitivity differential pressure gauges. These gauges are capable of detecting extremely small changes in differential pressure with a resolution of ∼0.02 Pa. This level of sensitivity allows meaningful pressure drop measurements across separate fuel components, even at low Reynolds numbers. The fuel assembly mock-ups were constructed from commercial fuel assembly structural components and stainless steel tubing that is within 0.6 pc of the outer diameter of actual fuel. The outer flow boundary in the BWR assembly bundle was defined by the walls of a prototypic canister. In the PWR assembly, the flow was confined by the walls of different stainless steel storage cells. Two of the PWR storage cell sizes represented dimensions spanning pool and cask cells available in industry. Pressure ports were installed along the length of the assemblies at locations corresponding to the entrance and exit of fuel components. Dry, ambient air was metered into the bottom of each assembly through a flow straightener. The geometries of the tube bundles in 17x17 PWR and 9x9 BWR fuel assemblies are fundamentally different. The PWR bundle has a larger flow area and incorporates more grid spacers compared to the BWR bundle. Additionally, eight of the 74 fuel rods in the 9x9

  6. Coupled code calculation of rod withdrawal at power accident

    Grgić, Davor, E-mail: davor.grgic@fer.hr [Faculty of Electrical Engineering and Computing, University of Zagreb, Unska 3, 10000 Zagreb (Croatia); Benčik, Vesna, E-mail: vesna.bencik@fer.hr [Faculty of Electrical Engineering and Computing, University of Zagreb, Unska 3, 10000 Zagreb (Croatia); Šadek, Siniša, E-mail: sinisa.sadek@fer.hr [Faculty of Electrical Engineering and Computing, University of Zagreb, Unska 3, 10000 Zagreb (Croatia)

    2013-08-15

    Highlights: ► Sensitivity calculations (withdrawal speed, initial power, secondary side influence) were performed for the rod withdrawal at power accident in PWR. ► Best estimate coupled RELAP5-PARCS code calculation was done, using COBRA code to model the core thermal-hydraulics. ► Specific modelling features included reactor vessel split model, explicit model of the RTD bypass and the overtemperature ΔT setpoint function. ► Average whole core values and the local hot spots were predicted. ► Local fuel centerline temperature and local DNBR were calculated using a COBRA-like model. ► Influence of the burnup on the fuel centerline temperature was studied. -- Abstract: The rod withdrawal at power (RWAP) accident is analyzed for NPP Krško as part of activity related to possible resistance temperature detectors (RTDs) bypass removal. The RWAP accident can be departure from nucleate boiling (DNB) or overpower limiting accident depending on initial power level and rate and amount of reactivity addition. In this paper we have analyzed the response of the plant in current configuration to RWAP for different withdrawal speeds and different initial power levels. By demonstrating adequacy of current protection system we can, in the next step, quantify the influence of change in narrow range coolant temperature measurement to available safety margins. The overtemperature ΔT setpoint and its relation to local DNBR values are in center of attention. The coupled RELAP5–PARCS code was used as the calculation tool with the provision to extend the calculation to local pin-by-pin COBRA subchannel calculation for selected state points derived from main coupled code results. In the first part of the calculation methodology, point kinetics calculation is performed using standalone RELAP5 to reproduce USAR results, and in the second part, more demanding coupled code calculation is introduced.

  7. Coupled code calculation of rod withdrawal at power accident

    Highlights: ► Sensitivity calculations (withdrawal speed, initial power, secondary side influence) were performed for the rod withdrawal at power accident in PWR. ► Best estimate coupled RELAP5-PARCS code calculation was done, using COBRA code to model the core thermal-hydraulics. ► Specific modelling features included reactor vessel split model, explicit model of the RTD bypass and the overtemperature ΔT setpoint function. ► Average whole core values and the local hot spots were predicted. ► Local fuel centerline temperature and local DNBR were calculated using a COBRA-like model. ► Influence of the burnup on the fuel centerline temperature was studied. -- Abstract: The rod withdrawal at power (RWAP) accident is analyzed for NPP Krško as part of activity related to possible resistance temperature detectors (RTDs) bypass removal. The RWAP accident can be departure from nucleate boiling (DNB) or overpower limiting accident depending on initial power level and rate and amount of reactivity addition. In this paper we have analyzed the response of the plant in current configuration to RWAP for different withdrawal speeds and different initial power levels. By demonstrating adequacy of current protection system we can, in the next step, quantify the influence of change in narrow range coolant temperature measurement to available safety margins. The overtemperature ΔT setpoint and its relation to local DNBR values are in center of attention. The coupled RELAP5–PARCS code was used as the calculation tool with the provision to extend the calculation to local pin-by-pin COBRA subchannel calculation for selected state points derived from main coupled code results. In the first part of the calculation methodology, point kinetics calculation is performed using standalone RELAP5 to reproduce USAR results, and in the second part, more demanding coupled code calculation is introduced

  8. Changes in 900 MW PWR alarm processing policy

    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

  9. Characterization of Factors affecting IASCC of PWR Core Internals

    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.

  10. Analysis and study on nuclear safety of Mitsubishi PWR

    Theme of safety analysis and study are changing to reflect the needs at the time. This paper introduces the overall aspects of transient and accident analysis performed and presents typical researches related to safety analysis for Mitsubishi PWR. (author)

  11. Fuel performance computer code simulation of steady-state and transient regimes of the stainless steel fuel rods

    The immediate cause of the accident at the Fukushima Daiichi nuclear plant in March 2011 was the meltdown of the reactor core. During this process, the zirconium cladding of the fuel reacts with water, producing a large amount of hydrogen. This hydrogen, combined with volatile radioactive materials leaked from the containment vessel and entered the building of the reactor, resulting in explosions. In the past, stainless steel was used as the coating in many pressurized water reactors (PWR) under irradiation and their performance was excellent, however, the stainless steel was replaced by a zirconium-based alloy as a coating material mainly due to its lower section shock-absorbing neutrons. Today, the stainless steel finish appears again as a possible solution for security issues related to the explosion and hydrogen production. The objective of this thesis is to discuss the performance under irradiation of fuel rods using stainless steel as a coating material. The results showed that stainless steel rods exhibit lower temperatures and higher fuel pellet width of the gap - coating the coated rods Zircaloy and this gap does not close during the irradiation. The thermal performance of the two fuel rods is very similar, and the penalty of increased absorption of neutrons due to the use of stainless steel can be offset by the combination of a small increase in the enrichment of U- 235 and changes in the size of the spacing between the fuel rods. (author)

  12. Hot Operation of FTL for PWR Fuels Irradiation

    Ahn, Sung Ho; Joung, Chang Yong; Lee, Jong Min; Park, Su Ki; Sim, Bong Sik; Ahn, Guk Hoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    Fuel Test Loop (FTL) in HANARO is the test facility which can conduct a fuel irradiation test with commercial NPPs' operating conditions such as their pressure, temperature, flow and water chemistry. The FTL is used for the irradiation test of PWR type or CNNDU type fuels. In this paper, the hot operation of FTL for irradiation test of PWR fuels is introduced. The experimental results show the excellence of operation performance

  13. Overview of US research related to PWR sump clogging

    In the framework of research of researches related to the PWR sump clogging in Usa, the author presents the history of GSI-191 (assessment of debris accumulation on PWR sump performance), the research to date (technical assessment, regulatory guide and evaluation guidance, model validation), the current and planned tests (chemical effect and calcium silicate tests, latent debris and downstream effect tests, integrated chemical effect tests, EPRI coatings study). (A.L.B.)

  14. Pressure-relieving devices and it's arrangement for PWR

    There are four types of PWR pressure-relieving devices: direct acting safety valve, pilot-operated pressure relief valve, power-operated pressure relief valve and safety valve with auxiliaries. The principle of operation, characteristics, arrangement of the pressure-relieving devices for PWR recently used at home and abroad, confidence of discharge, experience in service and developing trend of the devices are introduced. The first and second type of the devices are emphasised

  15. Delay stroke piston and rod for engine

    Booher, B.V.

    1995-03-09

    A reciprocating piston internal combustion engine comprises a cylinder having opposed ends, a piston reciprocably mounted in the cylinder, a connecting rod having a crank journal end and a piston journal end, the connecting rod connected to the piston at the piston journal end by means for first and second wrist pins spaced longitudinally along the rod, the first wrist pin journaled in a bore in the piston and in a slot in the piston rod, and the second wrist pin journaled in a bore in the piston rod and a longitudinal slot in the piston. (author)

  16. Snubber assembly for a control rod drive

    A snubber cartridge assembly is mounted to the nozzle of a control rod drive mechanism to insure that the snubber assembly will be located within the liquid filled section of a nuclear reactor vessel whenever the control rod drive is assembled thereto. The snubber assembly includes a piston mounted proximate to the control rod connecting end of the control rod drive leadscrew to allow the piston to travel within the liquid filled snubber cartridge and controllably exhaust liquid therefrom during a ''scram'' condition. The snubber cartridge provides three separate areas of increasing resistance to piston travel to insure a speedy but safe ''scram'' of the control rod into the reactor

  17. ELECTROMAGNETIC APPARATUS FOR MOVING A ROD

    Young, J.N.

    1958-04-22

    An electromagnetic apparatus for moving a rod-like member in small steps in either direction is described. The invention has particular application in the reactor field where the reactor control rods must be moved only a small distance and where the use of mechanical couplings is impractical due to the high- pressure seals required. A neutron-absorbing rod is mounted in a housing with gripping uaits that engage the rod, and coils for magnetizing the gripping units to make them grip, shift, and release the rod are located outside the housing.

  18. Analysis of Fly Fishing Rod Casting Dynamics

    Gang Wang; Norman Wereley

    2011-01-01

    An analysis of fly fishing rod casting dynamics was developed comprising of a nonlinear finite element representation of the composite fly rod and a lumped parameter model for the fly line. A nonlinear finite element model was used to analyze the transient response of the fly rod, in which fly rod responses were simulated for a forward casting stroke. The lumped parameter method was used to discretize the fly line system. Fly line motions were simulated during a cast based on fly rod tip resp...

  19. Rod ejection accident 3D-dynamic analysis in Almaraz NPP with RELAP5/PARCS V2.7 coupled codes

    The Rod Ejection Accident (REA) belongs to the Reactivity Initiated Accidents (RIA) category of accidents, and it is part of the licensing basis accident analyses required for pressurized water reactors (PWR). The REA consist of a rod ejection due to the failure of its driving mechanism. The evolution is driven by a continuous reactivity insertion. In previous works, we have analyzed this transient in Almaraz NPP at different power levels at the beginning of cycle (BOC) and at the end of cycle (EOC) using the coupled code RELAP5-MOD3.3/PARCS v2.7. In this work, we present the results of the REA analysis at hot zero power at BOC with all control rods inserted. In the thermal-hydraulic model used, each fuel assemblies has been modelled as an independent channel, for that, the RELAP5 source code has been modified and recompiled to accept this large number of channels. The neutronic nodal discretization consists of 157 x 24 active nodes, considering 13 different fuel elements with 291 neutronic compositions. The cross-sections sets are obtained from CASMO4-SIMULATE3 using the SIMTAB methodology developed in UPV. The transient departs from an initially critical core, being the withdrawal speed of the control rod a typical bounding value. These analyses will allow knowing more accurately the PWR real plant phenomenology in the RIA most limiting conditions, in this way the conclusions will be realistic. The aim is to improve the understanding of these accidents using advanced methods. (authors)

  20. Guide for rotating sucker rods

    Harrel, R.D.

    1986-11-04

    This patent describes an improved guide for use in a string of sucker rods rotated in a tubing string in a borehole, the sucker rods having threaded male ends, the guide comprising: an elongated upright solid cylindrical coupling body of external diameter less than the internal diameter of tubing in which it is to be used; a pair of spaced apart axle holders positioned in three recess; an axle received in each recess in the coupling body, the axis of each axle being parallel and spaced from the body longitudinal axis; a roller rotatably received on each axle, the periphery of each roller extending exteriorly of the external cylindrical surface of the coupling body; and means to retain each of the holders in the coupling body recess.