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

Directory of Open Access Journals (Sweden)

Surian Pinem

2014-01-01

Full Text Available A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of this work is to determine the accuracy of NODAL3 code in analysing the reactivity initiated accident due to the control rod ejection. The NEACRP PWR rod ejection cases are chosen since many organizations participated in the NEA project using various methods as well as approximations, so that, in addition to the reference solutions, the calculation results of NODAL3 code can also be compared to other codes’ results. The transient parameters to be verified are time of power peak, power peak, final power, final average Doppler temperature, maximum fuel temperature, and final coolant temperature. The results of NODAL3 code agree well with the PHANTHER reference solutions in 1993 and 1997 (revised. Comparison with other validated codes, DYN3D/R and ANCK, shows also a satisfactory agreement.

Comparison of rod-ejection transient calculations in hexagonal-Z geometry

International Nuclear Information System (INIS)

Knight, M.P.; Brohan, P.; Finnemann, H.; Huesken, J.

1995-01-01

This paper proposes a set of 3-dimensional benchmark rod ejection problems for a VVER reactor, based on the well-known NEACRP PWR rod-ejection problems defined by Siemens/KWU. Predictions for these benchmarks derived using three hexagonal-z nodal transient codes, the PANTHER code of Nuclear Electric, the HEXTIME code of Siemens/KWU, and the DYN3D code of FZ-Rossendorf are presented and compared

Uncertainties propagation in the framework of a Rod Ejection Accident modeling based on a multi-physics approach

Energy Technology Data Exchange (ETDEWEB)

Le Pallec, J. C.; Crouzet, N.; Bergeaud, V.; Delavaud, C. [CEA/DEN/DM2S, CEA/Saclay, 91191 Gif sur Yvette Cedex (France)

2012-07-01

The control of uncertainties in the field of reactor physics and their propagation in best-estimate modeling are a major issue in safety analysis. In this framework, the CEA develops a methodology to perform multi-physics simulations including uncertainties analysis. The present paper aims to present and apply this methodology for the analysis of an accidental situation such as REA (Rod Ejection Accident). This accident is characterized by a strong interaction between the different areas of the reactor physics (neutronic, fuel thermal and thermal hydraulic). The modeling is performed with CRONOS2 code. The uncertainties analysis has been conducted with the URANIE platform developed by the CEA: For each identified response from the modeling (output) and considering a set of key parameters with their uncertainties (input), a surrogate model in the form of a neural network has been produced. The set of neural networks is then used to carry out a sensitivity analysis which consists on a global variance analysis with the determination of the Sobol indices for all responses. The sensitivity indices are obtained for the input parameters by an approach based on the use of polynomial chaos. The present exercise helped to develop a methodological flow scheme, to consolidate the use of URANIE tool in the framework of parallel calculations. Finally, the use of polynomial chaos allowed computing high order sensitivity indices and thus highlighting and classifying the influence of identified uncertainties on each response of the analysis (single and interaction effects). (authors)

Ejected control rod and rods drop measurements during Mochovce startup physical tests

International Nuclear Information System (INIS)

Minarcin, Miroslav; Elko, Marek

1998-01-01

Paper deals with measurements of asymmetric reactivity insertion into the reactor core that were carried out during physical startup tests of Mochovce Unit 1 in June 1998. Control rods worth measurements with one and two rods s tucked in upper limit and worth measurement of one control rod from group 6 'ejected' from the reactor core are discussed. During the experiments neutron flux was measured by four ionisation chambers (three of them were placed symmetrically around the reactor core). Results of measurements and influence of asymmetric reactivity influence on ionisation chambers response are presented in the paper. (Authors)

Development of a DNBR evaluation method for the CEA ejection accident in SMART core

Energy Technology Data Exchange (ETDEWEB)

Hwang, Dae Hyun; Yoo, Y. J.; In, W. K.; Chang, M. H. [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-12-01

A methodology applicable to the analysis of the CEA ejection accident in SMART is developed for the evaluation of the fraction of fuel failure caused by DNB. The transient behavior of the core thermal-hydraulic conditions is calculated by the subchannel analysis code MATRA. The minimum DNBR during the accident is calculated by KRB-1 CHF correlation considering the 1/8 symmetry of hot assembly. The variation of hot assembly power during the accident is simulated by the LTC(Limiting transient Curve) which is determined from the analysis of power distribution data resulting from the three-dimensional core dynamics calculations. The initial condition of the accident is determined by considering LOC(Limiting Conditions for Operation) of SMART core. Two different methodologies for the evaluation of DNB failure rate are established; a deterministic method based on the DNB envelope, and a probabilistic method based on the DNB probability of each fuel rod. The methodology developed in this study is applied to the analysis of CEA ejection accident in the preliminary design core of SMART. As the result, the fractions of DNB fuel failure by the deterministic method and the probabilistic method are calculated as 38.7% and 7.8%, respectively. 16 refs., 16 figs., 5 tabs. (Author)

A three-dimensional pin-wise analysis for CEA ejection accident

Energy Technology Data Exchange (ETDEWEB)

Park, Guen-Tae; Park, Min-Ho; Park, Jin-Woo; Um, Kil-Sup; Choi, Tong-Soo [KEPCO NF, Daejeon (Korea, Republic of)

2016-10-15

The ejection of a control element assembly (CEA) with high reactivity worth causes the sudden insertion of reactivity into the core. Immediately after the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the doppler effect becomes important and turns the reactivity balance and power down to lower levels. The 3-D CEA ejection analysis methodology has been developed using the multi-dimensional code coupling system, CHASER, which couples three dimensional core neutron kinetics code ASTRA, subchannel analysis code THALES, and fuel performance analysis code FROST using message passing interface (MPI). This paper presents the pin-by-pin level analysis result with the 3-D CEA ejection analysis methodology using the CHASER. The pin-by-pin level analysis consists of DNBR, enthalpy and Pellet/Clad Mechanical Interaction (PCMI) analysis. All the evaluations are simulated for APR1400 plant loaded with PLUS7 fuel. In this paper, the pin-by-pin analysis using the multidimensional core transient code, CHASER, is presented with respect to enthalpy, DNBR and PCMI for APR1400 plant loaded with PLUS7 fuel. For the pin-by-pin enthalpy and DNBR analysis, the quarter core for HFP case or 15 - 20 assemblies around the most severe assembly for part powers or HZP cases are selected. And PCMI calculation is performed for all the rods in the whole core during a conservative time period. The pin-by-pin analysis results show that the regulatory guidelines of CEA ejection accident are satisfied.

A reactivity accidents simulation of the Fort Saint Vrain HTGR

International Nuclear Information System (INIS)

Fainer, Gerson

1980-01-01

A reactivity accidents analysis of the Fort Saint Vrain HTGR was made. The following accidents were analysed 1) A rod pair withdrawal accident during normal operation, 2) A rod pair ejection accident, 3) A rod pair withdrawal accident during startup operations at source levels and 4) Multiple rod pair withdrawal accident. All the simulations were performed by using the BLOOST-6 nuclear code The steady state reactor operation results obtained with the code were consistent with the design reactor data. The numerical analysis showed that all accidents - except the first one - cause particle failure. (author)

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

Directory of Open Access Journals (Sweden)

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.

Accident-tolerant control rod

International Nuclear Information System (INIS)

Ohta, Hirokazu; Sawabe, Takashi; Ogata, Takanari

2013-01-01

Boron carbide (B 4 C) and hafnium (Hf) metal are used for the neutron absorber materials of control rods in BWRs, and silver-indium-cadmium (Ag-In-Cd) alloy is used in PWRs. These materials are clad with stainless steel. The eutectic point of B 4 C and iron (Fe) is about 1150 deg. C and the melting point of Ag-In-Cd alloy is about 800 deg. C, which are lower than the temperature of zircaloy - steam reaction increases rapidly (∼1200 deg. C). Accordingly, it is possible that the control rods melt and collapse before the reactor core is significantly damaged in the case of severe accidents. Since the neutron absorber would be separated from the fuels, there is a risk of re-criticality, when pure water or seawater is injected for emergency cooling. In order to ensure sub-criticality and extend options of emergency cooling in the course of severe accidents, a concept of accident-tolerant control rod (ACT) has been derived. ACT utilises a new absorber material having the following properties: - higher neutron absorption than current control rod; - higher melting or eutectic temperature than 1200 deg. C where rapid zircaloy oxidation occurs; - high miscibility with molten fuel materials. The candidate of a new absorber material for ATC includes gadolinia (Gd 2 O 3 ), samaria (Sm 2 O 3 ), europia (Eu 2 O 3 ), dysprosia (Dy 2 O 3 ), hafnia (HfO 2 ). The melting point of these materials and the liquefaction temperature with Fe are higher than the rapid zircaloy oxidation temperature. ACT will not collapse before the core melt-down. After the core melt-down, the absorber material will be mixed with molten fuel material. The current absorber materials, such as B 4 C, Hf and Ag-In-Cd, are charged at the tip of ATC in which the neutron flux is high, and a new absorber material is charged in the low-flux region. This design could minimise the degradation of a new absorber material by the neutron absorption and the influence of ATC deployment on reactor control procedure. As a

Some elements of understanding about the cluster ejection accident in the EPR

International Nuclear Information System (INIS)

Vignon, Dominique

2010-01-01

The author answers to a publication made by an association (Sortir du Nucleaire) which is provided in appendix (some parts of this text are highlighted) and denounced risks associated with a cluster ejection accident in an EPR in relationship with steering modes which, according to this association, would be essentially related to an objective of economic profitability. The author first recalls some elements regarding the control and neutron stopping of pressurized water reactors. Then, after having outlined some specific aspects of the EPR design, he addresses the cluster ejection accident: safety approach and its application to this type of accident. He recalls the conclusions of studies of cluster ejection performed by EDF and AREVA, comments the consequences for the EPR power

Cobalt irradiation box ejection accident of ETRR-2

International Nuclear Information System (INIS)

El-Messiry, A.M.

2000-01-01

The new Egyptian test and research reactor number 2 ETRR-2, MTR type, is now under operational tests. It has a main central irradiation channel for the purpose of Co 60 isotope production with an intended rated capacity of 50000 Ci per year. The reactivity introduced in the reactor due to accidental ejection of the Co 60 irradiation box (CIB) should be discussed. This reactivity insertion accident (RIA) may be fast or slow with maximum reactivity worth 2.9428 $. The CIB may move with constant speed or variable acceleration according to its initial speed and the applied forces. This results in a linear, parabolic or sinusoidal motion, which in turn affects the reactivity insertion rate (RIR). The present work analyzes this type of perturbation during normal operating conditions: 22 MW full power and 1900 kg s -1 forced core cooling flow. The work serves as a part of the safety evaluation process applicable to similar MTR cores. The RIA code TRANSP20 is developed for this study. It simulates various types of RIR, fast or slow resulting from different CIB ejections. Scram signal due to power, period, inlet and outlet temperatures, or temperature difference is expected to activate the shutdown system. The work presents five case studies, two for fast ejection and three for slow. The transient behavior of the reactor during this is illustrated. The results show that the reactor can withstand slow ejection if the scram is available. However, for fast ejection the scram system does not prevent the clad temperature from exceeding safety limits. Recommendations to prevent or mitigate this accident are highlighted. (orig.)

Analysis of the NEACRP PWR rod ejection benchmark problems with DIF3D-K

International Nuclear Information System (INIS)

Kim, M.H.

1994-01-01

Analyses of the NEACRP PWR rod ejection transient benchmark problems with the DIF3D-K nodal kinetics code are presented. The DIF3D-K results are shown to be in generally good agreement with results obtained using other codes, in particular reference results previously generated with the PANTHER code. The sensitivity of the transient results to the DIF3D-K input parameters (such as time step size, radial and axial node sizes, and the mesh structure employed for fuel pin heat conduction calculation) are evaluated and discussed. In addition, the potential in reducing computational effort by application of the improved quasistatic scheme (IQS) to these rod ejection transients, which involve very significant flux shape changes and thermal-hydraulic feedback is evaluated

Finite-element solutions of the AER-2 rod ejection benchmark by CRONOS

International Nuclear Information System (INIS)

Kolev, N.P.; Lenain, R.; Fedon-Magnaud, C.

2001-01-01

The finite-element option in CRONOS was used to analyse the AER-2 rod-ejection benchmark for WWER-440. The objective is to obtain spatially converged solutions by means of node subdivision and approximation refinement. This paper presents the first phase of analysis dealing with the initial and just-ejected states used for calculation of the initial reactivity. Fine-mesh and extrapolated to zero mesh size solutions were obtained and verified by comparison to MAG code solutions. These differences provide potential for large deviations in the transient results and deserve further attention in reactor safety analysis (Authors)

Researches of WWER fuel rods behaviour under RIA accident conditions

International Nuclear Information System (INIS)

Nechaeva, O.; Medvedev, A.; Novikov, V.; Salatov, A.

2003-01-01

Unirradiated fuel rod and refabricated fuel rod tests in the BIGR as well as acceptance criteria proving absence of fragmentation and the settlement modeling of refabricated fuel rods thermomechanical behavior in the BIGR-tests using RAPTA-5 code are discussed in this paper. The behaviour of WWER type simulators with E110 and E635 cladding was researched at the BIGR reactor under power pulse conditions simulating reactivity initiated accident. The results of the tests in four variants of experimental conditions are submitted. The behaviour of 12 WWER type refabricated fuel rods was researched in the BIGR reactor under power pulse conditions simulating reactivity initiated accident: burnup 48 and 60 MWd/kgU, pulse width 3 ms, peak fuel enthalpy 115-190 cal/g. The program of future tests in the research reactor MIR with high burnup fuel rod (up to 70 MWd/kgU) under conditions simulating design RIA in WWER-1000 is presented

Examination of control rod ejection in WWER-440 type reactors at different circumstances using the code DYN3D

International Nuclear Information System (INIS)

Petoefi, G.; Aszodi, A.

2001-01-01

For nuclear reactors it is very important to examine the reactivity initiated transients caused by the ejection of a control rod. The event is found to be dependent on different thermal and neutronic parameters. In this paper the emphasis is laid on the effect of the power level at which the transient began and on the effect of the heat transfer coefficient measured in the gap between the fuel and the cladding. The most significant transients can be established by the ejection of the most effective control rod. So the first step is to determine the position of this rod. It was done by steady state calculations A calculation was carried out with all the rods inserted to the half level of the core, criticality was reached by adjusting the power level. Seven other calculations were made for each control rod at withdrawn position while the other six rods were inserted to the half plane of the core. From the results the most effective control rod could be determined.(authors)

Analysis Of Control Rod Ejection Of APR1400 By RELAP5

International Nuclear Information System (INIS)

Le Thi Thu; Hoang Minh Giang; Vo Thi Huong; Le Dai Dien

2011-01-01

This paper presents the analysis of Reactivity Induced Accident caused by ejection of a Control Element Assembly (CEA) from APR 1400 reactor vessel within 0.05 second. The initial condition were assumed as following: power level at 102%, delayed neutron fraction β = 412 pcm and CEA worth = 110 pcm. The analysis was simulated by RELAP5 code through two step: calculation of steady state and calculation of transient with initial condition mentioned as above. Some output results were presented with explanation: sequence of events corresponding to the time of the accident, the system behavior as power, reactivity feedback from fuel temperature changes (Doppler) as well as temperature, pressure, DNBR within 6 second of the accident. (author)

Fuel and control rod failure behavior during degraded core accidents

International Nuclear Information System (INIS)

Chung, K.S.

1984-01-01

As a part of the pretest and posttest analyses of Light Water Reactor Source Term Experiments (STEP) which are conducted in the Transient Reactor Test (TREAT) facility, this paper investigates the thermodynamic and material behaviors of nuclear fuel pins and control rods during severe core degradation accidents. A series of four STEP tests are being performed to simulate the characteristics of the power reactor accidents and investigate the behavior of fission product release during these accidents. To determine the release rate of the fission products from the fuel pins and the control rod materials, information concerning the timing of the clad failure and the thermodynamic conditions of the fuel pins and control rods are needed to be evaluated. Because the phase change involves a large latent heat and volume expansion, and the phase change is a direct cause of the clad failure, the understanding of the phase change phenomena, particularly information regarding how much of the fuel pin and control rod materials are melted are very important. A simple energy balance model is developed to calculate the temperature profile and melt front in various heat transfer media considering the effects of natural convection phenomena on the melting and freezing front behavior

Full-length fuel rod behavior under severe accident conditions

International Nuclear Information System (INIS)

Lombardo, N.J.; Lanning, D.D.; Panisko, F.E.

1992-12-01

This document presents an assessment of the severe accident phenomena observed from four Full-Length High-Temperature (FLHT) tests that were performed by the Pacific Northwest Laboratory (PNL) in the National Research Universal (NRU) reactor at Chalk River, Ontario, Canada. These tests were conducted for the US Nuclear Regulatory Commission (NRC) as part of the Severe Accident Research Program. The objectives of the test were to simulate conditions and provide information on the behavior of full-length fuel rods during hypothetical, small-break, loss-of-coolant severe accidents, in commercial light water reactors

Effects of B4C control rod degradation under severe accident

Energy Technology Data Exchange (ETDEWEB)

Seo, Si-Won; Park, Sang-Gil; Han, Sang-Ku [Atomic Creative Technology Co., Daejeon (Korea, Republic of)

2016-10-15

Boron carbide (B4C) is widely used as absorber material in western boiling water reactor (BWR), some PWR, EPR and Russian RBMK and VVERs. B4C oxidation is one of the important phenomena of in-vessel. In the present paper, the main results and knowledge gained regarding the B4C control rod degradation from above mentioned experiments are reviewed and arranged to inform its significance on the severe accident consequences. In this paper, the role of B4C control rod oxidation and the subsequent degradation on the severe accident consequences is reviewed with available literature and report of previous experimental program regarding the B4C oxidation. From this review, it seems that the contribution of this B4C oxidation on the accident progression to the further severe accident situation is not negligible. For the future work, the extensive experimental data interpretation will be performed to assess quantitatively the effect of B4C oxidation and degradation on the various postulated severe accident conditions.

A Basic Study on the Ejection of ICI Nozzle under Severe Accidents

Energy Technology Data Exchange (ETDEWEB)

Cho, Jong Rae; Bae, Ji Hoon; Bang, Kwang Hyun [Korea Maritime and Ocean University, Busan (Korea, Republic of); Park, Jong Woong [Dongguk University, Gyeongju (Korea, Republic of)

2016-05-15

Nozzle injection should be blocked because it affect to the environment if its melting core exposes outside. The purpose of this study is to carry out the thermos mechanical analysis due to debris relocation under severe accidents and to predict the nozzle ejection calculated considering the contact between the nozzle and lower head, and the supports of pipe cables. As a result of analyzing process of severe accidents, there was melting reaction between nozzle and the lower head. In this situation, we might predict the non-uniform contact region of nozzle hole of lower head and nozzle outside, delaying ejection of nozzles. But after melting, the average remaining length of the nozzle was 120mm and the maximum vertical displacement of lower nozzle near the weld is 3.3mm so there would be no nozzle this model, because the cable supports restrains the vertical displacement of nozzle.

Transient calculation performance of the MASTER code for control rod ejection problem

International Nuclear Information System (INIS)

Cho, B. O.; Joo, H. G.; Yoo, Y. J.; Park, S. Y.; Zee, S. Q.

1999-01-01

The accuracy and the effectiveness of the solution methods of the MASTER code for reactor transient problems were analyzed with a set of NEACRP PWR control rod ejection benchmark problems. A series of sensitivity study for the effects on the solution by the neutronic solution methods and the neutronic and thermal-hydraulic model parameters were thus investigated. The MASTER results were then compared with the reference PANTHER results. This indicates that the MASTER solution is sufficiently accurate and the computing time is fast enough for nuclear design application

Transient calculation performance of the MASTER code for control rod ejection problem

Energy Technology Data Exchange (ETDEWEB)

Cho, B. O.; Joo, H. G.; Yoo, Y. J.; Park, S. Y.; Zee, S. Q. [KAERI, Taejon (Korea, Republic of)

1999-10-01

The accuracy and the effectiveness of the solution methods of the MASTER code for reactor transient problems were analyzed with a set of NEACRP PWR control rod ejection benchmark problems. A series of sensitivity study for the effects on the solution by the neutronic solution methods and the neutronic and thermal-hydraulic model parameters were thus investigated. The MASTER results were then compared with the reference PANTHER results. This indicates that the MASTER solution is sufficiently accurate and the computing time is fast enough for nuclear design application.

Silver-indium-cadmium control rod behavior and aerosol formation in severe reactor accidents

International Nuclear Information System (INIS)

Petti, D.A.

1987-04-01

Silver-indium-cadmium (Ag-In-Cd) control rod behavior and aerosol formation in severe reactor accidents are examined in an attempt to improve the methodology used to estimate reactor accident source terms. Control rod behavior in both in-pile and out-of-pile experiments is reviewed. A mechanistic model named VAPOR is developed that calculates the downward relocation and simultaneous vaporization behavior of the Ag-In-Cd alloy expected after control rod failure in a severe reactor accident. VAPOR is used to predict the release of silver, indium, and cadmium vapors expected in the Power Burst Facility (PBF) Severe Fuel Damage (SFD) 1-4 experiment. In addition, a sensitivity study is performed. Although cadmium is found to be the most volatile constituent of the alloy, all of the calculations predict that the rapid relocation of the alloy down to cooler portions of the core results in a small release for all three control rod alloy vapors. Potential aerosol formation mechanisms in a severe reactor accident are reviewed. Specifically, models for homogeneous, ion-induced, and heterogeneous nucleation are investigated. These models are applied to silver, cadmium, and CsI to examine the nucleation behavior of these three potential aerosol sources in a severe reactor accident and to illustrate the competition among these mechanisms for vapor depletion. The results indicate that aerosol formation in a severe reactor accident occurs in three stages. In the first stage, ion-induced nucleation causes aerosol generation. During the second stage, ion-induced and heterogeneous nucleation operates as competing pathways for gas-to-particle conversion until sufficient aerosol surface area is generated. In the third stage, ion-induced nucleation ceases; and heterogeneous nucleation becomes the dominant mechanism of gas-to-particle conversion until equilibrium is reached

Control rod drop accident analysis for the mixed core project in Ling Ao NPS

International Nuclear Information System (INIS)

Zhang Shishun; Zhou Zhou; Xiao Min

2004-01-01

AFA-2G assemblies in Ling Ao NPS (LNPS) have been replaced gradually by AFA-3G assemblies from cycle 2 and subsequent cycles. the enrichment of the fuels will be increased from 3.2% to 3.7% from cycle 3 in Ling Ao. Therefore, the study of ling Ao mixed core and increased enrichment have been performed since 2001. Lots of accidents need to be re-analyzed in Ling Ao NPS in order to verify its safety requirements for the new fuel management. Control rod drop accident for LNPS was re-analyzed in 2001 in frame of FRAMATOME ANP analytical methodology. The analytical codes used in the accident analysis include SCIENCE, ESPADON, CINEMA, CANTAL and FLICA III. The control rod drop accident analysis is performed with respect to the 10 reference cycles of the generic fuel management design for Ling Ao mixed core and increased enrichment study. The pre-drop FδH for the first transition cycles and other cycles are 1.52 and 1.55, respectively. For detected dropped rod configurations, the negative flux rate protection system actuates a reactor trip. For the non-detected dropped rod configurations, the minimum DNBR values have been evaluated with conservative analysis methodology and assumptions and the DNBR fuel design limit is respected the analytical results shows that, for all the non-detected dropped rod configurations, the minimum DNB margin is about 2% which occurs in AFA-2G fuel assembly in the first transition cycle. (author)

Accident analysis for PRC-II reactor

International Nuclear Information System (INIS)

Wei Yongren; Tang Gang; Wu Qing; Lu Yili; Liu Zhifeng

1997-12-01

The computer codes, calculation models, transient results, sensitivity research, design improvement, and safety evaluation used in accident analysis for PRC-II Reactor (The Second Pulsed Reactor in China) are introduced. PRC-II Reactor is built in big populous city, so the public pay close attention to reactor safety. Consequently, Some hypothetical accidents are analyzed. They include an uncontrolled control rod withdrawal at rated power, a pulse rod ejection at rated power, and loss of coolant accident. Calculation model which completely depict the principle and process for each accident is established and the relevant analysis code is developed. This work also includes comprehensive computing and analyzing transients for each accident of PRC-II Reactor; the influences in the reactor safety of all kind of sensitive parameters; evaluating the function of engineered safety feature. The measures to alleviate the consequence of accident are suggested and taken in the construction design of PRC-II Reactor. The properties of reactor safety are comprehensively evaluated. A new advanced calculation model (True Core Uncovered Model) of LOCA of PRC-II Reactor and the relevant code (MCRLOCA) are first put forward

Computation of reactor control rod drop time under accident conditions

International Nuclear Information System (INIS)

Dou Yikang; Yao Weida; Yang Renan; Jiang Nanyan

1998-01-01

The computational method of reactor control rod drop time under accident conditions lies mainly in establishing forced vibration equations for the components under action of outside forces on control rod driven line and motion equation for the control rod moving in vertical direction. The above two kinds of equations are connected by considering the impact effects between control rod and its outside components. Finite difference method is adopted to make discretization of the vibration equations and Wilson-θ method is applied to deal with the time history problem. The non-linearity caused by impact is iteratively treated with modified Newton method. Some experimental results are used to validate the validity and reliability of the computational method. Theoretical and experimental testing problems show that the computer program based on the computational method is applicable and reliable. The program can act as an effective tool of design by analysis and safety analysis for the relevant components

Analysis of the rod drop accident for Angra-1

International Nuclear Information System (INIS)

Veloso, M.A.; Atayde, P.A.

1989-01-01

The aim of this work is to present a rod drop accident analysis for the third cycle of the Angra-1 nuclear power plant operating in the automatic control mode. In this analysis all possible configurations for dropped rods caused by a single failure in the controller circuits have been considered. The dropped rod worths, power distributions and excore detector tilts were determined by using the Siemens/KWU neutronic code system, in particular the MEDIUM2, PINPOW and DETILT codes. The transient behaviour of the plant during the rod drop event was simulated with the SACI2/MOD0 code, developed at CDTN. Determinations related to the DNBR design limit were conducted by utilizing the CDTN PANTERA-1P subchannel code. The transient analysis indicated that for dropped rod worths greater than about 425 pcm reactor trip from negative neutron flux rate will take place independently of core conditions. In the range from 0 to 425 pcm large power overshoots may occur as a consequence of the automatic control system action. The magnitude of the maximum power peaking during the event increases with the dropped rod worth, as far as the control bank is able to compensate the initial reactivity decrease. Thermal-hydraulic evaluations carried out with the PANTERA-1P code show that for all the relevant dropped rod worths the minimum DNBR will remain above a limit value of 1.365. Even if this conservative limit is met, the calculated nuclear power peaking factors, F N AH , will be at least 6% higher than the allowable F N AH -values. Therefore, the DNBR design margin will be preserved at the event of rod drop. (author)

The safety feature of hydraulic driving system of control rod for 200 MW nuclear heating reactor

International Nuclear Information System (INIS)

Chi Zongbo; Wu Yuanqiang

1997-01-01

The hydraulic driving system of control rod is used as control rod drive mechanism in 200 MW nuclear heating reactor. Design of this system is based on passive system, integrating drive and guide of control rod. The author analyzes the inherent safety and the design safety of this system, with mechanism of control rod not ejecting when the pressure of pressure vessel is lost, and calculating result of core not exposing when the amount of coolant is drained by broken pipe. The results indicate that this system has good safety feature, and assures reactor safety under any accident conditions, providing important technology support for 200 MW nuclear heating reactor with inherent safety feature

SSYST-1. A computer code system to analyse the fuel rod behaviour during a loss of coolant accident

International Nuclear Information System (INIS)

Gulden, W.

1977-08-01

The modules of the SSYST program system allow the detailed analysis of an LWR fuel rod in the course of a postulated loss-of-coolant accident. They provide a tool for considering the interaction between the heat conduction in the fuel rod, heat transfer in the gap, fuel and cladding tube deformation, pressure in the coolant, as well as thermal and fluid dynamics in the cooling channel and for calculating the time and location of ballooning and rod failure, respectively. They can be used both to precalculate the behaviour of fuel rods during LWR accidents and in support of the design of experiments. Depending on the problem to be solved, the individual modules can be easily combined. (orig.) [de

A filter system for steam-gas mixture ejections from under a nuclear reactor containment following a severe accident

International Nuclear Information System (INIS)

Dulepov, Ju. N.; Sharygin, L. M.; Tretjakov, S. Ja.; Shtin, A.P.; Glushko, V. V.; Babenko, E. A.; Kurakov, Ju. A.

1997-01-01

In this paper newly built NPPs obligatory incorporate a containment having a filter system for removing radioactive materials ejections under severe accidents including nuclear fuel melting is described. The system prevents a containment failure and provides ejected radioactive materials decontamination to permissible levels. The physical-chemical and chemical characteristics of Termoxid-58 sorbent (TiO 5 based sorbent) are presented

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

International Nuclear Information System (INIS)

Sengpiel, W.

1980-12-01

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) [de

Silver-indium-cadmium control rod behaviour during a severe reactor accident

International Nuclear Information System (INIS)

Bowsher, B.R.; Jenkins, R.A.; Nichols, A.L.; Rowe, N.A.; Simpson, J.A.H.

1986-04-01

An alloy of silver, indium and cadmium is commonly used as control rod material in pressurised water reactors (PWRs). The behaviour of this alloy has been studied in a series of experiments using an induction furnace to achieve temperatures up to 1900K. The aerosols released from overheated clad and unclad control rod samples have been characterised in both steam and inert atmospheres. Mass balance experiments have been undertaken to determine the distribution of the control rod alloy constituents following rupture of the cladding, and this work has been supported by thermogravimetric studies of silver-indium mixtures. Metallographic studies were also undertaken to assess the failure mode of the stainless steel cladding and the interaction of the molten alloy with Zircaloy. The results of this work are discussed in terms of aerosol/vapour behaviour during severe reactor accidents. (author)

Analysis of PWR control rod ejection accident with the coupled code system SKETCH-INS/TRACE by incorporating pin power reconstruction model

International Nuclear Information System (INIS)

Nakajima, T.; Sakai, T.

2010-01-01

The pin power reconstruction model was incorporated in the 3-D nodal kinetics code SKETCH-INS in order to produce accurate calculation of three-dimensional pin power distributions throughout the reactor core. In order to verify the employed pin power reconstruction model, the PWR MOX/UO_2 core transient benchmark problem was analyzed with the coupled code system SKETCH-INS/TRACE by incorporating the model and the influence of pin power reconstruction model was studied. SKETCH-INS pin power distributions for 3 benchmark problems were compared with the PARCS solutions which were provided by the host organisation of the benchmark. SKETCH-INS results were in good agreement with the PARCS results. The capability of employed pin power reconstruction model was confirmed through the analysis of benchmark problems. A PWR control rod ejection benchmark problem was analyzed with the coupled code system SKETCH-INS/ TRACE by incorporating the pin power reconstruction model. The influence of pin power reconstruction model was studied by comparing with the result of conventional node averaged flux model. The results indicate that the pin power reconstruction model has significant effect on the pin powers during transient and hence on the fuel enthalpy

Control assembly ejection accident analysis for WWER-440 (Armenian NPP)

International Nuclear Information System (INIS)

Bznuni, S.; Malakyan, Ts.; Amirjanyan, A.; Ghasabyan, L.

2007-01-01

Control Assembly ejection in WWER-440 initiated by the loss of integrity of the Control Assemblies drive housing has been analyzed. This event causes a very rapid reactivity insertion to the core and small break LOCA which potentially could lead to rapid power increase and redistribution of heat release in the core resulting in a fuel, cladding and coolant temperature rise; primary pressure increase, radiological consequences due to loss of primary coolant and potential loss of cladding integrity and fuel disintegration (if applicable). Methodology of the analysis is based on conservative assumptions as well as on deterministic approach for selection of functioning logic of systems and equipment's to maximize reactor core power and minimize power decreasing reactivity feedback. Computational analyses were performed by 3D kinetics PARCS-RELAP coupled code. WWER-440 fuel cross-section libraries, diffusion coefficients and kinetics parameters were calculated by HELOS code. In this paper analysis of accident for Hot Full Power was presented. Results of analysis show that ANPP WWER-440 reactor design meets acceptance criteria prescribed for RIA type design based accidents (Authors)

Laboratory simulation of rod-to-rod mechanical interactions during postulated loss-of-coolant accidents in a PWR involving cladding oxidation

International Nuclear Information System (INIS)

Hindle, E.D.; Haste, T.J.; Harrison, W.R.

1987-01-01

Creep deformation of Zircaloy cladding in postulated PWR loss-of-coolant accidents may lead to rod-to-rod mechanical interactions. Tests have been performed in the electrically heated FOURSQUARE rig at 750 0 C and 850 0 C in steam to investigate this effect. Conservatisms inherent in a simple 'square with rounded corners' coolant channel blockage model have been quantified; about 5-10% flow area may remain even at strains which in ideal circumstances would give total blockage. Reduction of average burst strains produced by an oxide layer (up to 13 μm) has been demonstrated, resulting from strain concentration at oxide cracks. (author)

Improved point-kinetics model for the BWR control rod drop accident

International Nuclear Information System (INIS)

Neogy, P.; Wakabayashi, T.; Carew, J.F.

1985-01-01

A simple prescription to account for spatial feedback weighting effects in RDA (rod drop accident) point-kinetics analyses has been derived and tested. The point-kinetics feedback model is linear in the core peaking factor, F/sub Q/, and in the core average void fraction and fuel temperature. Comparison with detailed spatial kinetics analyses indicates that the improved point-kinetics model provides an accurate description of the BWR RDA

Study of a criticality accident involving fuel rods and water outside a power reactor

International Nuclear Information System (INIS)

Beloeil, L.

2000-01-01

It is possible to imagine highly unlikely but numerous accidental situations where fuel rods come into contact with water under conditions close to atmospheric values. This work is devoted to modelling and simulation of first instants of the power excursion that may result from such configurations. We show that void effect is a preponderant feedback for most severe accidents. The formation of a vapour film around the rods is put forward and confirmed with the help of experimental transients using electrical heating. We propose then a vapour/liquid flow model able to reproduce void fraction evolution. The vapour film is treated as a compressible medium. Conservation balance equations are solved on a moving mesh with a two-dimensional scheme and boundary conditions taking notice of interfacial phenomena and axial escape possibility. Movements of the liquid phase are modelled through a non-stationary integral equation and a dissipative term suited to the particular geometry of this flow. The penetration of energy into the liquid is also calculated. Thus, the coupling of aerodynamic and hydrodynamic modules gives results in excellent agreement with experiments. Next, neutronic phenomena into the fuel pellet, their feedback effects and the distribution of power through the rod are numerically translated. For each developed module, validation tests are provided. Then, it is possible to simulate the first seconds of the whole criticality accident. Even if this calculation tool is only a way of study as a first approach, performed simulations are proving coherent with reported data on recorded accidents. (author)

A methodology for the evaluation of fuel rod failures under transportation accidents

International Nuclear Information System (INIS)

Rashid, J.Y.R.; Machiels, A.J.

2004-01-01

Recent studies on long-term behavior of high-burnup spent fuel have shown that under normal conditions of stor-age, challenges to cladding integrity from various postulated damage mechanisms, such as delayed hydride crack-ing, stress-corrosion cracking and long-term creep, would not lead to any significant safety concerns during dry storage, and regulatory rules have subsequently been established to ensure that a compatible level of safety is maintained. However, similar safety assurances for spent fuel transportation have not yet been developed, and further studies are currently being conducted to evaluate the conditions under which transportation-related safety issues can be resolved. One of the issues presently under evaluation is the ability and the extent of the fuel as-semblies to maintain non-reconfigured geometry during transportation accidents. This evaluation may determine whether, or not, the shielding, confinement, and criticality safety evaluations can be performed assuming initial fuel assembly geometries. The degree to which spent fuel re-configuration could occur during a transportation accident would depend to a large degree on the number of fuel rod failures and the type and geometry of the failure modes. Such information can only be developed analytically, as there is no direct experimental data that can provide guidance on the level of damage that can be expected. To this end, the paper focuses on the development of a modeling and analysis methodology that deals with this general problem on a generic basis. First consideration is given to defining acci-dent loading that is equivalent to the bounding, although analytically intractable, hypothetical transportation acci-dent of a 9-meter drop onto essentially unyielding surface, which is effectively a condition for impact-limiters de-sign. Second, an analytically robust material constitutive model, an essential element in a successful structural analysis, is required. A material behavior model

Total Monte-Carlo method applied to the assessment of uncertainties in a reactivity-initiated accident

Energy Technology Data Exchange (ETDEWEB)

Cruz, D.F. da; Rochman, D.; Koning, A.J. [Nuclear Research and Consultancy Group NRG, Petten (Netherlands)

2014-07-01

The Total Monte-Carlo (TMC) method has been applied extensively since 2008 to propagate the uncertainties in nuclear data for reactor parameters and fuel inventory, and for several types of advanced nuclear systems. The analyses have been performed considering different levels of complexity, ranging from a single fuel rod to a full 3-D reactor core at steady-state. The current work applies the TMC method for a full 3-D pressurized water reactor core model under steady-state and transient conditions, considering thermal-hydraulic feedback. As a transient scenario the study focused on a reactivity-initiated accident, namely a control rod ejection accident initiated by a mechanical failure of the control rod drive mechanism. The uncertainties on the main reactor parameters due to variations in nuclear data for the isotopes {sup 235},{sup 238}U, {sup 239}Pu and thermal scattering data for {sup 1}H in water were quantified. (author)

3D analysis of the reactivity insertion accident in VVER-1000

Energy Technology Data Exchange (ETDEWEB)

Abdullayev, A. M.; Zhukov, A. I.; Slyeptsov, S. M. [NSC Kharkov Inst. for Physics and Technology, 1, Akademicheskaya Str., Kharkov 61108 (Ukraine)

2012-07-01

Fuel parameters such as peak enthalpy and temperature during rod ejection accident are calculated. The calculations are performed by 3D neutron kinetics code NESTLE and 3D thermal-hydraulic code VIPRE-W. Both hot zero power and hot full power cases were studied for an equilibrium cycle with Westinghouse hex fuel in VVER-1000. It is shown that the use of 3D methodology can significantly increase safety margins for current criteria and met future criteria. (authors)

Control rod drive shaft latch

International Nuclear Information System (INIS)

Thorp, A.G. II.

1976-01-01

A latch mechanism is operated by differential pressure on a piston to engage the drive shaft for a control rod in a nuclear reactor, thereby preventing the control rod from being ejected from the reactor in case of failure of the control rod drive mechanism housing which is subjected to the internal pressure in the reactor vessel. 6 claims, 4 drawing figures

SSYST. A code system to analyze LWR fuel rod behavior under accident conditions

International Nuclear Information System (INIS)

Gulden, W.; Meyder, R.; Borgwaldt, H.

1982-01-01

SSYST (Safety SYSTem) is a modular system to analyze the behavior of light water reactor fuel rods and fuel rod simulators under accident conditions. It has been developed in close cooperation between Kernforschungszentrum Karlsruhe (KfK) and the Institut fuer Kerntechnik und Energiewandlung (IKE), University Stuttgart, under contract of Projekt Nukleare Sicherheit (PNS) at KfK. Although originally aimed at single rod analysis, features are available to calculate effects such as blockage ratios of bundles and wholes cores. A number of inpile and out-of-pile experiments were used to assess the system. Main differences versus codes like FRAP-T with similar applications are (1) an open-ended modular code organisation, (2) availability of modules of different sophistication levels for the same physical processes, and (3) a preference for simple models, wherever possible. The first feature makes SSYST a very flexible tool, easily adapted to changing requirements; the second enables the user to select computational models adequate to the significance of the physical process. This leads together with the third feature to short execution times. The analysis of transient rod behavior under LOCA boundary conditions e.g. takes 2 mins cpu-time (IBM-3033), so that extensive parametric studies become possible

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

Energy Technology Data Exchange (ETDEWEB)

Purcell, P.C. [BNFL International Transport, Spent Fuel Services (United Kingdom); Dallongeville, M. [COGEMA Logistics (AREVA Group) (France)

2004-07-01

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.

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

International Nuclear Information System (INIS)

Purcell, P.C.; Dallongeville, M.

2004-01-01

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

Self-cleaning threaded rod spinneret for high-efficiency needleless electrospinning

Science.gov (United States)

Zheng, Gaofeng; Jiang, Jiaxin; Wang, Xiang; Li, Wenwang; Zhong, Weizheng; Guo, Shumin

2018-07-01

High-efficiency production of nanofibers is the key to the application of electrospinning technology. This work focuses on multi-jet electrospinning, in which a threaded rod electrode is utilized as the needless spinneret to achieve high-efficiency production of nanofibers. A slipper block, which fits into and moves through the threaded rod, is designed to transfer polymer solution evenly to the surface of the rod spinneret. The relative motion between the slipper block and the threaded rod electrode promotes the instable fluctuation of the solution surface, thus the rotation of threaded rod electrode decreases the critical voltage for the initial multi-jet ejection and the diameter of nanofibers. The residual solution on the surface of threaded rod is cleaned up by the moving slipper block, showing a great self-cleaning ability, which ensures the stable multi-jet ejection and increases the productivity of nanofibers. Each thread of the threaded rod electrode serves as an independent spinneret, which enhances the electric field strength and constrains the position of the Taylor cone, resulting in high productivity of uniform nanofibers. The diameter of nanofibers decreases with the increase of threaded rod rotation speed, and the productivity increases with the solution flow rate. The rotation of electrode provides an excess force for the ejection of charged jets, which also contributes to the high-efficiency production of nanofibers. The maximum productivity of nanofibers from the threaded rod spinneret is 5-6 g/h, about 250-300 times as high as that from the single-needle spinneret. The self-cleaning threaded rod spinneret is an effective way to realize continuous multi-jet electrospinning, which promotes industrial applications of uniform nanofibrous membrane.

Linking of FRAP-T, FRAPCON and RELAP-4 codes for transient analysis and accidents of light water reactors fuel rods

International Nuclear Information System (INIS)

Marra Neto, A.; Silva, A.T. e; Sabundjian, G.; Freitas, R.L.; Neves Conti, T. das.

1991-09-01

The computer codes FRAP-T, FRAPCON and RELAP-4 have been linked for the fuel rod behavior analysis under transients and hypothetical accidents in light water reactors. The results calculated by thermal hydraulic code RELAP-4 give input in file format into the transient fuel analysis code FRAP-T. If the effect of fuel burnup is taken into account, the fuel performance code FRAPCON should provide the initial steady state data for thhe transient analysis. With the thermal hydraulic boundary conditions provided by RELAP-4 (MOD3), FRAP-T6 is used to analyse pressurized water reactor fuel rod behavior during the blowdown phase under large break loss of coolant accident conditions. Two cases have been analysed: without and with initialization from FRAPCON-2 steady state data. (author)

Experimental data report for Test TS-1 Reactivity Initiated Accident Test in NSRR with pre-irradiated BWR fuel rod

International Nuclear Information System (INIS)

Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Horiki, Ohichiro; Yamahara, Takeshi; Ichihashi, Yoshinori; Kikuchi, Teruo

1992-01-01

This report presents experimental data for Test TS-1 which was the first in a series of tests, simulating Reactivity Initiated Accident (RIA) conditions using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in October, 1989. Test fuel rod used in the Test TS-1 was a short-sized BWR (7 x 7) type rod which was fabricated from a commercial rod provided from Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79 % and burnup of 21.3 GWd/t (bundle average). Pulse irradiation was performed at a condition of stagnant water cooling, atmospheric pressure and ambient temperature using a newly developed double container-type capsule. Energy deposition of the rod in this test was evaluated to be about 61 cal/g·fuel (55 cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, fuel burnup measurements, transient behavior of the test rod during pulse irradiation and results of post pulse irradiation examinations are contained in this report. (author)

Experimental data report for Test TS-2 reactivity initiated accident test in NSRR with pre-irradiated BWR fuel rod

International Nuclear Information System (INIS)

Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Kobayashi, Shinsho; Yamahara, Takeshi; Sukegawa, Tomohide; Kikuchi, Teruo

1993-02-01

This report presents experimental data for Test TS-2 which was the second test in a series of Reactivity Initiated Accident (RIA) condition test using pre-irradiated BWR fuel rods, performed at the Nuclear Safety Research Reactor (NSRR) in February, 1990. Test fuel rod used in the Test TS-2 was a short sized BWR (7x7) type rod which was fabricated from a commercial rod irradiated at Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79% and a burnup of 21.3Gwd/tU (bundle average). A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The energy deposition of the fuel rod in this test was evaluated to be 72±5cal/g·fuel (66±5cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and, results of pre and post pulse irradiation examinations are described in this report. (author)

75 FR 39975 - Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards...

Science.gov (United States)

2010-07-13

.... At normal operating pressures, leakage from Primary Water Stress Corrosion Cracking (PWSCC) below 16...) will not occur. No leakage factor will be applied to the Locked Rotor or Control Rod Ejection due to... Line Break evaluation Locked Rotor evaluation Control Rod Ejection evaluation Loss of Coolant Accident...

Analyzing the BWR rod drop accident in high-burnup cores

International Nuclear Information System (INIS)

Diamond, D.J.; Neymotin, L.; Kohut, P.

1995-01-01

This study was undertaken for the US Nuclear Regulatory Commission to determine the fuel enthalpy during a rod drop accident (RDA) for cores with high burnup fuel. The calculations were done with the RAMONA-4B code which models the core with 3-dimensional neutron kinetics and multiple parallel coolant channels. The calculations were done with a model for a BWR/4 with fuel bundles having burnups up to 30 GWd/t and also with a model with bundle burnups to 60 GWd/t. This paper also discusses potential sources of uncertainty in calculations with high burnup fuel. One source is the ''rim'' effect which is the extra large peaking of the power distribution at the surface of the pellet. This increases the uncertainty in reactor physics and heat conduction models that assume that the energy deposition has a less peaked spatial distribution. Two other sources of uncertainty are the result of the delayed neutron fraction decreasing with burnup and the positive moderator temperature feedback increasing with burnup. Since these effects tend to increase the severity of the event, an RDA calculation for high burnup fuel will underpredict the fuel enthalpy if the effects are not properly taken into account. Other sources of uncertainty that are important come from the initial conditions chosen for the RDA. This includes the initial control rod pattern as well as the initial thermal-hydraulic conditions

Status and results of the theoretical and experimental investigations on the LWR fuel rod behavior under accident conditions

International Nuclear Information System (INIS)

Bocek, M.; Hofmann, P.; Leistikow, S.; Class, G.; Meyder, R.; Raff, S.; Erbacher, F.; Hofmann, G.; Ihle, P.; Karb, E.; Fiege, A.

1978-09-01

In this report the status of knowledge is described which has been gathered up to the end of 1977 of the LWR fuel rod behavior in loss-of-coolant accidents. The majority of results indicated have been derived from studies on the fuel rod behavior performed within the framework of the Nuclear Safety Project (PNS); partly, also the results of cooperating research establishments and fm international exchange of experience are referred to. The report has been subdivided into two complete parts: Part I provides a survey of the most significant results of the theoretical and experimental research projects on fuel rod behavior. Part II describes by detailed individual presentations the status as well as the results with respect to the major central subjects. (orig.) 891 RW 892 AP [de

Lumped-parameter fuel rod model for rapid thermal transients

International Nuclear Information System (INIS)

Perkins, K.R.; Ramshaw, J.D.

1975-07-01

The thermal behavior of fuel rods during simulated accident conditions is extremely sensitive to the heat transfer coefficient which is, in turn, very sensitive to the cladding surface temperature and the fluid conditions. The development of a semianalytical, lumped-parameter fuel rod model which is intended to provide accurate calculations, in a minimum amount of computer time, of the thermal response of fuel rods during a simulated loss-of-coolant accident is described. The results show good agreement with calculations from a comprehensive fuel-rod code (FRAP-T) currently in use at Aerojet Nuclear Company

Microstructural examination of fuel rods subjected to a simulated large-break loss of coolant accident in reactor

International Nuclear Information System (INIS)

Garlick, A.

1985-01-01

A series of tests has been conducted in the National Research Universal (NRU) reactor, Chalk River, Canada, to investigate the behaviour of full-length 32-rod PWR fuel bundles during a simulated large-break loss of coolant accident (LOCA). In one of these tests (MT-3), 12 central rods were pre-pressurized in order to evaluate the ballooning and rupture of cladding in the Zircaloy high-α/α+β temperature region. All 12 rods ruptured after experiencing < 90% diametral strain but there was no suggestion of coplanar blockage. Post-irradiation examination was carried out on cross-sections of cladding from selected rods to determine the aximuthal distribution of wall thinning along the ballooned regions. These data are assessed to check whether they are consistent with a mechanism in which fuel stack eccentricity generates temperature gradients around the ballooning cladding and leads to premature rupture during a LOCA. After anodizing, the cladding microstructures were examined for the presence of prior-beta phase that would indicate the α/α+β transformation temperature (1078K) had been exceeded. These results were compared with isothermal annealing test data on unirradiated cladding from the same manufacturing batch

Boiling water reactor radiation shielded Control Rod Drive Housing Supports

Energy Technology Data Exchange (ETDEWEB)

Baversten, B.; Linden, M.J. [ABB Combustion Engineering Nuclear Operations, Windsor, CT (United States)

1995-03-01

The Control Rod Drive (CRD) mechanisms are located in the area below the reactor vessel in a Boiling Water Reactor (BWR). Specifically, these CRDs are located between the bottom of the reactor vessel and above an interlocking structure of steel bars and rods, herein identified as CRD Housing Supports. The CRD Housing Supports are designed to limit the travel of a Control Rod and Control Rod Drive in the event that the CRD vessel attachement went to fail, allowing the CRD to be ejected from the vessel. By limiting the travel of the ejected CRD, the supports prevent a nuclear overpower excursion that could occur as a result of the ejected CRD. The Housing Support structure must be disassembled in order to remove CRDs for replacement or maintenance. The disassembly task can require a significant amount of outage time and personnel radiation exposure dependent on the number and location of the CRDs to be changed out. This paper presents a way to minimize personal radiation exposure through the re-design of the Housing Support structure. The following paragraphs also delineate a method of avoiding the awkward, manual, handling of the structure under the reactor vessel during a CRD change out.

Experimental data report for test TS-3 Reactivity Initiated Accident test in the NSRR with pre-irradiated BWR fuel rod

International Nuclear Information System (INIS)

Nakamura, Takehiko; Yoshinaga, Makio; Fujishiro, Toshio; Kobayashi, Shinsho; Yamahara, Takeshi; Sukegawa, Tomohide; Kikuchi, Teruo; Sobajima, Makoto.

1993-09-01

This report presents experimental data for Test TS-3 which was the third test in a series of Reactivity Initiated Accident (RIA) tests using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in September, 1990. Test fuel rod used in the Test TS-3 was a short-sized BWR (7 x 7) type rod which was re-fabricated from a commercial rod irradiated in the Tsuruga Unit 1 power reactor of Japan Atomic Power Co. The fuel had an initial enrichment of 2.79 % and a burnup of 26 Gwd/tU. A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The energy deposition of the fuel rod in this test was evaluated to be 94 ± 4 cal/g · fuel (88 ± 4 cal/g · fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and results of pre-pulse and post-pulse irradiation examinations are described in this report. (author)

CTF/DYN3D multi-scale coupled simulation of a rod ejection transient on the NURESIM platform

Directory of Open Access Journals (Sweden)

Yann Périn

2017-09-01

Full Text Available In the framework of the EU funded project NURESAFE, the subchannel code CTF and the neutronics code DYN3D were integrated and coupled on the NURESIM platform. The developments achieved during this 3-year project include assembly-level and pin-by-pin multiphysics thermal hydraulics/neutron kinetics coupling. In order to test this coupling, a PWR rod ejection transient was simulated on a MOX/UOX minicore. The transient is simulated using two different models of the minicore. In the first simulation, both codes model the core with an assembly-wise resolution. In the second simulation, a pin-by-pin fuel-centered model is used in CTF for the central assembly, and a pin power reconstruction method is applied in DYN3D. The analysis shows the influence of the different models on global parameters, such as the power and the average fuel temperature, but also on local parameters such as the maximum fuel temperature.

Modeling of continuous withdrawal and falling out of CPS control rods accident, using QUABOX/CUBBOX-HYCA code

International Nuclear Information System (INIS)

Bubelis, E.; Pabarcius, R.; Tonkunas, A.

2003-01-01

At present, at the Ignalina NPP the process of a wider use of the new uranium-erbium fuel of higher saturation and the manual control rods of new design is going on. These actions are directed to reducing the reactor control and protection system (CPS) cooling circuit voiding effect and to improving the technical and economical reactor operation parameters. Continuous withdrawal and falling out of CPS control rods lead to the reactivity and power changes in the reactor core. Therefore, important for safety is the evaluation of the CPS ability to compensate for the resulting excess reactivity in the reactor core, having the changed core loading conditions during such accidents. This article presents the calculation results of the continuous withdrawal and falling out of CPS control rods for the specific reactor core conditions of the Ignalina NPP Unit 2, i.e. during its operation on the maximum allowed power level of 4200 MW. The German code QUABOX/CUBBOX-HYCA with the improved CPS logic was used for the simulation of the above-mentioned transients. (author)

Calculation of hydrogen and oxygen uptake in fuel rod cladding during severe accidents using the integral diffusion method -- Preliminary design report

International Nuclear Information System (INIS)

Siefken, L.J.

1999-01-01

Preliminary designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert's law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the proposed models on the calculated behavior of fuel rods in severe accident conditions. This report is a revision and reissue of the report entitled; ''Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents.''

A Comparative analysis for control rod drop accident in WH and CE type

International Nuclear Information System (INIS)

Yang, Chang-Keun; Kim, Yo-Han; Ha, Jun-Sang

2008-01-01

In Korea, the nuclear industries such as fuel manufacturer, the architect engineer and the utility, have been using the methodologies and codes of vendors, such as Westinghouse(WH), Combustion Engineering, for the safety analyses of nuclear power plants. Consequently the industries have kept up the many organizations to operate the methodologies and to maintain the codes for each vendor. It may occur difficulty to improve the safety analyses efficiency and technology related. So, the necessity another of methodologies and code systems applicable to Non- LOCA, beyond design basis accident and performance analyses for all types of pressurized water reactor(PWR) has been raised. Due to the above reason, the Korea Electric Power Research Institute(KEPRI) had decided to develop the new safety analysis code system for Korea Standard Nuclear Power Plants in Korea. As the first requirement, the best-estimate codes were required for applicable wider application area and realistic behavior prediction of power plants with various and sophisticated functions. After the investigation for few candidates, RETRAN-3D has been chosen as a system analysis code. As a part of the feasibility estimation for the methodology and code system, CRD(Control Rod Drop) accident which an event of Non-LOCA accidents for Uljin units 3 and 4 and Yonggwang 1 and 2 was selected to verify the feasibility of the methodology using the RETRAN-3D. And the results were compared with those mentioned in the final safety analysis reports (FSARs) of the plants

Substitute safety rods: Physics design and NTG calibration

International Nuclear Information System (INIS)

Baumann, N.P.

1991-07-01

Under certain assumed accident conditions, an SRS reactor may loose most of its bulk moderator while maintaining flow to fuel assemblies. If this occurs immediately after operation at power, components normally dependent on convective heat transfer to the moderator will heat up with the possibility of melting that component. One component at risk is the safety rod. Tests have shown that the current cadmium safety rod, which contains aluminum as well as cadmium, can fail at temperatures only slightly in excess of 500 deg C. Computations indicate that such temperatures can be reached with operating powers well below the 50% power limit now imposed by other accident scenarios. Safety rod melting would thus establish a new lower operating limit. A substitute safety rod that could tolerate much higher temperatures would eliminate this limit. This memorandum details the physics characteristics of a suitable replacement rod. 7 refs

Stability and failure analysis of steering tie-rod

Science.gov (United States)

Jiang, GongFeng; Zhang, YiLiang; Xu, XueDong; Ding, DaWei

2008-11-01

A new car in operation of only 8,000 km, because of malfunction, resulting in lost control and rammed into the edge of the road, and then the basic vehicle scrapped. According to the investigation of the site, it was found that the tie-rod of the car had been broken. For the subjective analysis of the accident and identifying the true causes of rupture of the tierod, a series of studies, from the angle of theory to experiment on the bended broken tie-rod, were conducted. The mechanical model was established; the stability of the defective tie-rod was simulated based on ANSYS software. Meanwhile, the process of the accident was simulated considering the effect of destabilization of different vehicle speed and direction of the impact. Simultaneously, macro graphic test, chemical composition analysis, microstructure analysis and SEM analysis of the fracture were implemented. The results showed that: 1) the toughness of the tie-rod is at a normal level, but there is some previous flaws. One quarter of the fracture surface has been cracked before the accident. However, there is no relationship between the flaw and this incident. The direct cause is the dynamic instability leading to the large deformation of impact loading. 2) The declining safety factor of the tie-rod greatly due to the previous flaws; the result of numerical simulation shows that previous flaw is the vital factor of structure instability, on the basis of the comparison of critical loads of the accident tie-rod and normal. The critical load can decrease by 51.3% when the initial defect increases 19.54% on the cross-sectional area, which meets the Theory of Koiter.

Controlled tungsten melting and droplet ejection studies in ASDEX Upgrade

International Nuclear Information System (INIS)

Krieger, K; Lunt, T; Dux, R; Janzer, A; Müller, H W; Potzel, S; Pütterich, T; Yang, Z

2011-01-01

Tungsten rods of 1×1×3 mm 3 were exposed in single H-mode discharges at the outer divertor target plate of ASDEX Upgrade using the divertor manipulator system. Melting of the W rod at a pre-defined time was induced by moving the initially far away outer strike point close to the W-rod position. Visible light emissions of both the W pin and consecutively ejected W droplets were recorded by two fast cameras with crossed viewing cones. The time evolution of the local W source at the pin location was measured by spectroscopic observation of the WI line emission at 400.9 nm and compared to the subsequent increase of tungsten concentration in the confined plasma derived from tungsten vacuum UV line emission. Combining these measurements with the total amount of released tungsten due to the pin melt events and ejected droplets allowed us to derive an estimate of the screening factor for this type of tungsten source. The resulting values of the tungsten divertor retention in the range 10-20 agree with those found in previous studies using a W source of sublimated W(CO) 6 vapour at the same exposure location. Ejected droplets were found to be always accelerated in the general direction of the plasma flow, attributed to friction forces and to rocket forces. Furthermore, the vertically inclined target plates cause the droplets, which are repelled by the target plate surface potential due to their electric charge, to move upwards against gravity due to the centrifugal force component parallel to the target plate.

Fuel pin behaviour under conditions of control rod withdrawal accident in CABRI-2 experiments

International Nuclear Information System (INIS)

Papin, Joelle; Lemoine, Francette; Sato, Ikken; Struwe, Dankward; Pfrang, Werner

1994-01-01

Simulation of the control rod withdrawal accident has been performed in the international CABRI-2 experimental programme. The tests realized with industrial pins led to clarification of the influence of the pellet design and have shown the important role of fission products on the solid fuel swelling which promotes early pin failure with solid fuel pellet. With annular pellet design, large fuel swelling combined to low smear density leads to degradation of fuel thermal conductivity and thus reduces power to melt. However, the high margin to deterministic failure is confirmed with hollow pellets. Improvements of the modelling were necessary to describe such behaviours in computer codes as SAS-4A, PAPAS-2S and PHYSURAC. (author)

Safety rod driving device

International Nuclear Information System (INIS)

Murakami, Kiyonobu; Kurosaki, Akira.

1988-01-01

Purpose: To rapidly insert safety rods for a criticality experiment device into a reactor core container to stop the criticality reaction thereby prevent reactivity accidents. Constitution: A cylinder device having a safety rod as a cylinder rod attached with a piston at one end is constituted. The piston is elevated by pressurized air and attracted and fixed by an electromagnet which is a stationary device disposed at the upper portion of the cylinder. If the current supply to the electromagnet is disconnected, the safety rod constituting the cylinder rod is fallen together with the piston to the lower portion of the cylinder. Since the cylinder rod driving device has neither electrical motor nor driving screw as in the conventional device, necessary space can be reduced and the weight is decreased. In addition, since the inside of the nuclear reactor can easily be shielded completely from the external atmosphere, leakage of radioactive materials can be prevented. (Horiuchi, T.)

Influence of initial conditions on rod behaviour during boiling crisis phase following a reactivity initiated accident

International Nuclear Information System (INIS)

Georgenthum, V.; Sugiyama, T.

2010-01-01

In the frame of their research programs on high burn-up fuel safety, the French Institute for Radioprotection and Nuclear Safety (IRSN) and the Japan Atomic Energy Agency (JAEA) performed a large set of tests devoted to the study of PWR fuel rod behavior during Reactivity Initiated Accident (RIA) respectively in the CABRI reactor and in the NSRR reactor. The reactor test conditions are different in terms of coolant nature, temperature and pressure. In the CABRI reactor, tests were performed until now with sodium coolant at 280 Celsius degrees and 3 bar. In the NSRR reactor most of the tests were performed with stagnant water at 20 C. degrees and atmospheric pressure but recently a new high temperature high pressure capsule has been developed which allows to performed tests at up to 280 Celsius degrees and 70 bar. The paper discusses the influence of test conditions on rod behaviour during boiling phase, based on tests results and SCANAIR code calculations. The study shows that when the boiling crisis is reached, the initial inner and outer rod pressure have an essential impact on the clad straining and possible ballooning. The analysis of the different test conditions makes it possible to discriminate the influence of initial conditions on the different phases of the transient and is useful for modelling and code development. (authors)

Control Rod Drive Mechanism Installed in the Internal of Reactor Pressure Vessel

Energy Technology Data Exchange (ETDEWEB)

Choi, M. H.; Choi, S.; Park, J. S.; Lee, J. S.; Kim, D. O.; Hur, N. S.; Hur, H.; Yu, J. Y

2008-09-15

This report describes the review results and important technologies related to the in-vessel type control rod drive mechanism. Generally, most of the CRDMs used in the PWR are attached outside of the reactor pressure vessel, and the pernetration of the vessel head can not avoid. However, in-vessel type CRDMs, which are installed inside the reactor vessel, can eliminate the possibility of rod ejection accidents and the penetration of the vessel head, and provide a compact design of the reactor vessel and containment. There are two kinds of in-vessel type CRDM concerning the driving force-driven by a driving motor and by a hydraulic force. Motor driven CRDMs have been mainly investigated in Japan(MRX, IMR, DRX, next generation BWR etc.), and developed the key components such as a canned motor, an integrated rod position indicator, a separating ball-nut and a ball bearing that can operate under the water conditions of a high temperature and pressure. The concept of hydraulically driven CRDMs have been first reported by KWU and Siemens for KWU 200 reactor, and Argentina(CAREM) and China(NHR-5, NHR-200) have been developed the internal CRDM with the piston and cylinder of slightly different geometries. These systems are driven by the hydraulic force which is produced by pumps outside of the reactor vessel and transmitted through a pipe penetrating the reactor vessel, and needs complicated control and piping systems including pumps, valves and pipes etc.. IRIS has been recently decided the internal CRDMs as the reference design, and an analytical and experimental investigations of the hydraulic drive concept are performed by POLIMI in Italy. Also, a small French company, MP98 has been developed a new type of control rods, called 'liquid control rods', where reactivity is controlled by the movement of a liquid absorber in a manometer type device.

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

International Nuclear Information System (INIS)

Gadelha, J.A.M.

1982-10-01

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

Advantages of Westinghouse BWR control rod drop accidents methodology utilizing integrated POLCA-T code

International Nuclear Information System (INIS)

Panayotov, Dobromir

2008-01-01

The paper focuses on the activities pursued by Westinghouse in the development and licensing of POLCA-T code Control Rod Drop Accident (CRDA) Methodology. The comprehensive CRDA methodology that utilizes PHOENIX4/POLCA7/POLCA-T calculation chain foresees complete cycle-specific analysis. The methodology consists of determination of candidates of control rods (CR) that could cause a significant reactivity excursion if dropped throughout the entire fuel cycle, selection of limiting initial conditions for CRDA transient simulation and transient simulation itself. The Westinghouse methodology utilizes state-of-the-art methods. Unnecessary conservatisms in the methodology have been avoided to allow the accurate prediction of margin to design bases. This is mainly achieved by using the POLCA-T code for dynamic CRDA evaluations. The code belongs to the same calculation chain that is used for core design. Thus the very same reactor, core, cycle and fuel data base is used. This allows also reducing the uncertainties of input data and parameters that determine the energy deposition in the fuel. Uncertainty treatment, very selective use of conservatisms, selection of the initial conditions for limiting case analyses, incorporation into POLCA-T code models of the licensed fuel performance code are also among the means of performing realistic CRDA transient analyses. (author)

Study of a criticality accident involving fuel rods and water outside a power reactor; Etude d'un accident de criticite mettant en presence des crayons combustibles et de l'eau hors reacteur de puissance

Energy Technology Data Exchange (ETDEWEB)

Beloeil, L

2000-05-30

It is possible to imagine highly unlikely but numerous accidental situations where fuel rods come into contact with water under conditions close to atmospheric values. This work is devoted to modelling and simulation of first instants of the power excursion that may result from such configurations. We show that void effect is a preponderant feedback for most severe accidents. The formation of a vapour film around the rods is put forward and confirmed with the help of experimental transients using electrical heating. We propose then a vapour/liquid flow model able to reproduce void fraction evolution. The vapour film is treated as a compressible medium. Conservation balance equations are solved on a moving mesh with a two-dimensional scheme and boundary conditions taking notice of interfacial phenomena and axial escape possibility. Movements of the liquid phase are modelled through a non-stationary integral equation and a dissipative term suited to the particular geometry of this flow. The penetration of energy into the liquid is also calculated. Thus, the coupling of aerodynamic and hydrodynamic modules gives results in excellent agreement with experiments. Next, neutronic phenomena into the fuel pellet, their feedback effects and the distribution of power through the rod are numerically translated. For each developed module, validation tests are provided. Then, it is possible to simulate the first seconds of the whole criticality accident. Even if this calculation tool is only a way of study as a first approach, performed simulations are proving coherent with reported data on recorded accidents. (author)

Tensile and burst tests in support of the cadmium safety rod failure evaluation

International Nuclear Information System (INIS)

Thomas, J.K.

1992-02-01

The reactor safety rods may be subjected to high temperatures due to gamma heating after the core coolant level has dropped during the ECS phase of hypothetical LOCA event. Accordingly, an experimental safety rod testing subtask was established as part of a task to address the response of reactor core components to this accident. This report discusses confirmatory separate effects tests conducted to support the evaluation of failures observed in the safety rod thermal tests. As part of the failure evaluation, the potential for liquid metal embrittlement (LME) of the safety rod cladding by cadmium (Cd) -- aluminum (Al) solutions was examined. Based on the test conditions, literature data, and U-Bend tests, its was concluded that the SS304 safety rod cladding would not be subject to LME by liquid Cd-Al solutions under conditions relevant to the safety rod thermal tests or gamma heating accident. To confirm this conclusion, tensile tests on SS304 specimens were performed in both air and liquid Cd-Al solutions with the range of strain rates, temperatures, and loading conditions spanning the range relevant to the safety rod thermal tests and gamma heating accident

Radiation dose distributions due to sudden ejection of cobalt device

International Nuclear Information System (INIS)

Abdelhady, Amr

2016-01-01

The evaluation of the radiation dose during accident in a nuclear reactor is of great concern from the viewpoint of safety. One of important accident must be analyzed and may be occurred in open pool type reactor is the rejection of cobalt device. The study is evaluating the dose rate levels resulting from upset withdrawal of co device especially the radiation dose received by the operator in the control room. Study of indirect radiation exposure to the environment due to skyshine effect is also taken into consideration in order to evaluate the radiation dose levels around the reactor during the ejection trip. Microshield, SHLDUTIL, and MCSky codes were used in this study to calculate the radiation dose profiles during cobalt device ejection trip inside and outside the reactor building. - Highlights: • This study aims to calculate the dose rate profiles after cobalt device ejection from open-pool-type reactor core. • MicroShield code was used to evaluate the dose rates inside the reactor control room. • McSKY code was used to evaluate the dose rates outside the reactor building. • The calculated dose rates for workers are higher than the permissible limits after 18 s from device ejection.

Simulation of reactivity-initiated accident transients on UO2-M5® fuel rods with ALCYONE V1.4 fuel performance code

Directory of Open Access Journals (Sweden)

Isabelle Guénot-Delahaie

2018-03-01

Full Text Available The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs, power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs. As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on PWR-UO2 fuel rods with advanced claddings such as M5® under “low pressure–low temperature” or “high pressure–high temperature” water coolant conditions.This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on UO2-M5® fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE—starting from base irradiation conditions it itself computes—is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur.Areas of improvement are finally discussed with a view to simulating and

Behavior of defective LWR-type fuel rods irradiated under postulated accident conditions

International Nuclear Information System (INIS)

Hobbins, R.R.; Croucher, D.W.; Seiffert, S.L.; Cook, B.A.; Kerwin, D.K.; Mehner, A.S.; Ploger, S.A.

1979-05-01

The irradiation experiments reported here have been conducted by the Thermal Fuels Behavior Program of EG and G Idaho, Inc., for the United States Nuclear Regulatory Commission in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. Five of the rods were irradiated in PCM tests and one in a LOC test. During these tests, the six rods lost cladding integrity prior to or during the transient phase of the test due to either manufacturing defects or intentional rod design and operation. Of the five defective rods tested under PCM conditions, one (Rod IE-008, Test IE-1) had a hydride rupture below the region of the rod, which was in film boiling during the transient; two (Rod A-0021, Test PCM-3 and Rod IE-019, Test IE-5) contained defects (a pin hole and a small axial crack, respectively) within the film boiling zone; and two (Rod 201-1, Test PCM-1 and Rod 205-8, Test PCM-5) failed by cladding embrittlement within the film boiling zone. Rod 312-3 was waterlogged before being subjected to LOC conditions in Test LLR-3

Survivability rate among pilots in case of ejection

Directory of Open Access Journals (Sweden)

Alexandru GHEORGHIU

2015-06-01

Full Text Available The current paper presents a statistical analysis of a recent research made by the author [1], showing the factors causing the accidents that happened in Romanian Air Force from 1952 to 2014. Also the decision of ejection is analyzed. The study contains 225 events: 110 catastrophes and 115 accidents. 280 fighter pilots and 235 aircraft were involved in this analysis. The below information is a personal one and does not reflect an official position of the Ministry of National Defence.

Simulation of nuclear fuel rods by using process computer-controlled power for indirect electrically heated rods

International Nuclear Information System (INIS)

Malang, S.

1975-11-01

An investigation was carried out to determine how the simulation of nuclear fuel rods with indirect electrically heated rods could be improved by use of a computer to control the electrical power during a loss-of-coolant accident (LOCA). To aid in the experiment, a new version of the HETRAP code was developed which simulates a LOCA with heater rod power controlled by a computer that adjusts rod power during a blowdown to minimize the difference in heat flux of the fuel and heater rods. Results show that without computer control of heater rod power, only the part of a blowdown up to the time when the heat transfer mode changes from nucleate boiling to transition or film boiling can be simulated well and then only for short times. With computer control, the surface heat flux and temperature of an electrically heated rod can be made nearly identical to that of a reactor fuel rod with the same cooling conditions during much of the LOCA. A small process control computer can be used to achieve close simulation of a nuclear fuel rod with an indirect electrically heated rod

A space-time analysis of PWR rod ejection accident

International Nuclear Information System (INIS)

Aoki, K.; Kanezashi, M.; Kato, M.; Nishimura, Y.; Satoh, T.

1985-01-01

This paper presents an efficient and reliable method to perform V-Q loss minimization control. A method proposed here is suitable for a decentralized control system (a hierarchy control system), taking full advantage of the dispatching organization and system structures. This is also a practical method, taking computational requirements and transmitted information into consideration. A feasible type, decomposition technique of quadratic programming underlies our method. Numerical results show that execution time, memory requirements, and convergence property are satisfactory

A Comparative analysis for control rod drop accident in RETRAN DNB and CETOP DNB Model

International Nuclear Information System (INIS)

Yang, Chang Keun; Kim, Yo Han; Ha, Sang Jun

2009-01-01

In Korea, the nuclear industries such as fuel manufacturer, the architect engineer and the utility, have been using the methodologies and codes of vendors, such as Westinghouse(WH), Combustion Engineering, for the safety analyses of nuclear power plants. Consequently the industries have kept up the many organizations to operate the methodologies and to maintain the codes for each vendor. It may occur difficulty to improve the safety analyses efficiency and technology related. So, the necessity another of methodologies and code systems applicable to Non- LOCA, beyond design basis accident and performance analyses for all types of pressurized water reactor(PWR) has been raised. Due to the above reason, the Korea Electric Power Research Institute(KEPRI) had decided to develop the new safety analysis code system for Korea Standard Nuclear Power Plants in Korea. As the first requirement, the best-estimate codes were required for applicable wider application area and realistic behavior prediction of power plants with various and sophisticated functions. After the investigation for few candidates, RETRAN-3D has been chosen as a system analysis code. As a part of the feasibility estimation for the methodology and code system, CRD(Control Rod Drop) accident which an event of Non-LOCA accidents for Uljin units 3 and 4 and Yonggwang 1 and 2 was selected to verify the feasibility of the methodology using the RETRAN-3D. In this paper, RETRAN DNB Model and CETOP DNB Model were analyzed by using comparative method

Preliminary analysis of control rod accidents in the CRCN-R1 multipurpose reactor core of Recife in Brazil

International Nuclear Information System (INIS)

Souza dos Santos, Rubens; Rubens Maiorino, Jose

1999-01-01

The paper shows some results of the neutronic accident analyses arisen by uncontrolled control rod withdrawal, based on the Conceptual Project of the CRCN-R1 MultiPurpose Reactor of Recife. In that reactor, a project of the CNEN/Brazil, under the leadership of the IPEN/Sao Paulo, is verified the thermal hydraulic limits in the reactor core during transients that simulate startup and power operation accidents. It has utilized a computer program that solved the kinetic equations based on multigroup diffusion theory, in our case we have used 4 energy groups, Two-Dimensional X-Y in the space, and 6 groups of delayed neutrons. A simple model of feedback is admitted in the capture and scattering macroscopic cross sections, in the fuel regions, temperature and coolant densities dependents. Based on those models, the results demonstrated that the reactor exhibits good degree of safety. (author)

Qualification of ARROTTA code for LWR accident analysis

International Nuclear Information System (INIS)

Huang, P.-H.; Peng, K.Y.; Lin, W.-C.; Wu, J.-Y.

2004-01-01

This paper presents the qualification efforts performed by TPC and INER for the 3-D spatial kinetics code ARROTTA for LWR core transient analysis. TPC and INER started a joint 5 year project in 1989 to establish independent capabilities to perform reload design and transient analysis utilizing state-of-the-art computer programs. As part of the effort, the ARROTTA code was chosen to perform multi-dimensional kinetics calculations such as rod ejection for PWR and rod drop for BWR. To qualify ARROTTA for analysis of FSAR licensing basis core transients, ARROTTA has been benchmarked for the static core analysis against plant measured data and SIMULATE-3 predictions, and for the kinetic analysis against available benchmark problems. The static calculations compared include critical boron concentration, core power distribution, and control rod worth. The results indicated that ARROTTA predictions match very well with plant measured data and SIMULATE-3 predictions. The kinetic benchmark problems validated include NEACRP rod ejection problem, 3-D LMW LWR rod withdrawal/insertion problem, and 3-D LRA BWR transient benchmark problem. The results indicate that ARROTTA's accuracy and stability are excellent as compared to other space-time kinetics codes. It is therefore concluded that ARROTTA provides accurate predictions for multi-dimensional core transient for LWRs. (author)

Modeling and Experimental Tests on the Hydraulically Driven Control Rod option for IRIS Reactor

International Nuclear Information System (INIS)

Cammi, Antonio; Ricotti, Marco E.; Vitulo, Alessia

2004-01-01

The adoption of Internal Control Rod Drive Mechanisms (ICRDMs) represents a valuable alternative to classical, external CRDMs based on electro-magnetic devices, as adopted in current PWRs. The advantages on the safety features of the reactor are apparent: inherent elimination of the Rod Ejection accidents and of possible concerns about the vessel head penetrations. A further positive feedback on the design is the reduction of the primary system overall dimensions. Within the frame of the ICRDM concepts, the Hydraulically Driven Control Rod solution is investigated as a possible option for the IRIS integral reactor. After a brief comparison of the solutions currently proposed for integral reactors, the configuration of the Hydraulic Control Rod device for IRIS, made up by an external movable piston and an internal fixed cylinder, is described. A description of the whole control system is reported as well. Particular attention is devoted to the Control Rod profile characterization, performed by means of a Computational Fluid Dynamics (CFD) analysis. The investigation of the system behavior has been carried out, including the dynamic equilibrium and its stability properties, the withdrawal and insertion step movement and the sensitivity study on command time periods. A suitable dynamic model has been set up for the mentioned purposes: the models corresponding to the various Control Rod system devices have been written in an Object-Oriented language (Modelica), thus allowing an easy implementation of such a system into the simulator for the whole reactor. Finally, a preliminary low pressure, low temperature, reduced length experimental facility has been built. Tests on HDCR stability and operational transients have been performed. The results are compared with the dynamic system model and CFD simulation model, showing good agreement between simulations and experimental data. During these preliminary tests, the control system performed correctly, allowing stable dynamic

Cooling of safety rods in the Savannah River K Reactor during the gamma heating phase of a postulated loss-of-coolant accident

International Nuclear Information System (INIS)

Pasamehmetoglu, K.O.; Unal, C.; Motley, F.E.; Rodriguez, S.B.

1992-01-01

This paper documents the heat-transfer analysis for the safety rod placed in a perforated guide tube during the gamma heating phase of a large-break loss of coolant accident in Savannah River K-reactor. The cooling mechanisms are natural convection to air and radiation to the surrounding structures. The limiting component is the guide tube. The guide tube is shown to remain coolable below its thermal limit for the anticipated reactor powers unless it is contacted by the hotter safety rod. Sample calculations are performed for various contact scenarios, and the results are reported within the paper. The results indicate that the most limiting contact scenario results when the safety rod heats up to its maximum temperature while remaining concentric in the guide tube and then contacts the guide tube. The worse contact location appears to be in line with the slugs-cladding contact and in between the rows of holes in the guide tube

Radiation dose distributions due to sudden ejection of cobalt device.

Science.gov (United States)

Abdelhady, Amr

2016-09-01

The evaluation of the radiation dose during accident in a nuclear reactor is of great concern from the viewpoint of safety. One of important accident must be analyzed and may be occurred in open pool type reactor is the rejection of cobalt device. The study is evaluating the dose rate levels resulting from upset withdrawal of co device especially the radiation dose received by the operator in the control room. Study of indirect radiation exposure to the environment due to skyshine effect is also taken into consideration in order to evaluate the radiation dose levels around the reactor during the ejection trip. Microshield, SHLDUTIL, and MCSky codes were used in this study to calculate the radiation dose profiles during cobalt device ejection trip inside and outside the reactor building. Copyright © 2016 Elsevier Ltd. All rights reserved.

Assessment of Core Failure Limits for Light Water Reactor Fuel under Reactivity Initiated Accidents

International Nuclear Information System (INIS)

Jernkvist, Lars Olof; Massih, Ali R.

2004-12-01

Core failure limits for high-burnup light water reactor UO 2 fuel rods, subjected to postulated reactivity initiated accidents (RIAs), are here assessed by use of best-estimate computational methods. The considered RIAs are the hot zero power rod ejection accident (HZP REA) in pressurized water reactors and the cold zero power control rod drop accident (CZP CRDA) in boiling water reactors. Burnup dependent core failure limits for these events are established by calculating the fuel radial average enthalpy connected with incipient fuel pellet melting for fuel burnups in the range of 30 to 70 MWd/kgU. The postulated HZP REA and CZP CRDA result in lower enthalpies for pellet melting than RIAs that take place at rated power. Consequently, the enthalpy thresholds presented here are lower bounds to RIAs at rated power. The calculations are performed with best-estimate models, which are applied in the FRAPCON-3.2 and SCANAIR-3.2 computer codes. Based on the results of three-dimensional core kinetics analyses, the considered power transients are simulated by a Gaussian pulse shape, with a fixed width of either 25 ms (REA) or 45 ms (CRDA). Notwithstanding the differences in postulated accident scenarios between the REA and the CRDA, the calculated core failure limits for these two events are similar. The calculated enthalpy thresholds for fuel pellet melting decrease gradually with fuel burnup, from approximately 960 J/gUO 2 at 30 MWd/kgU to 810 J/gUO 2 at 70 MWd/kgU. The decline is due to depression of the UO 2 melting temperature with increasing burnup, in combination with burnup related changes to the radial power distribution within the fuel pellets. The presented fuel enthalpy thresholds for incipient UO 2 melting provide best-estimate core failure limits for low- and intermediate-burnup fuel. However, pulse reactor tests on high-burnup fuel rods indicate that the accumulation of gaseous fission products within the pellets may lead to fuel dispersal into the coolant at

Post-accident core coolability of light water reactors

International Nuclear Information System (INIS)

Michio, I.; Teruo, I.; Tomio, Y.; Tsutao, H.

1983-01-01

A study on post-accident core coolability of LWR is discussed based on the practical fuel failure behavior experienced in NSRR, PBF, PNS and others. The fuel failure behavior at LOCA, RIA and PCM conditions are reviewed, and seven types of fuel failure modes are extracted as the basic failure mechanism at accident conditions. These are: cladding melt or brittle failure, molten UO 2 failure, high temperature cladding burst, low temperature cladding burst, failure due to swelling of molten UO 2 , failure due to cracks of embrittled cladding for irradiated fuel rods, and TMI-2 core failure. The post-accident core coolability at each failure mode is discussed. The fuel failures caused actual flow blockage problems. A characteristic which is common among these types is that the fuel rods are in the conditions violating the present safety criteria for accidents, and UO 2 pellets are in melting or near melting hot conditions when the fuel rods failed

Analysis of Rod Withdrawal at Power (RWAP) Accident using ATHLET Mod 2.2 Cycle A and RELAP5/mod 3.3 Codes

International Nuclear Information System (INIS)

Bencik, V.; Cavlina, N.; Grgic, D.

2012-01-01

The system code ATHLET is being developed at Gesellschaft fuer Anlagen-und Reaktorsicherheit (GRS) in Germany. In 1996, the NPP Krsko (NEK) input deck for ATHLET Mod 1.1 Cycle C has been developed at Faculty of Electrical Engineering (FER), University of Zagreb. The input deck was tested by analyzing the realistic plant event 'Main Steam Isolation Valve Closure' and the results were assessed against the measured data. The input deck was established before plant modernization that took place in 2000 and included the power uprate and SG replacement. The released ATHLET version (Mod 2.2 Cycle A) is now being available at FER Zagreb. Accordingly, the NEK input deck for ATHLET Mod 2.2 Cycle A has been developed. A completely new input deck has been created taking into account the large number of changes due to power uprate and SG replacement as well as taking advantage of developmental work on NEK data base performed at FER. The new NEK input deck for ATHLET code has been tested by analyzing the Rod Withdrawal Power (RWAP) accident and the results were assessed against the analysis performed by RELAP5/mod 3.3 code. The RWAP accident can be either Departure from Nucleate Boiling (DNB) ratio or overpower limiting accident depending on initial power and reactivity insertion rate. Since the automatic rod control system is assumed unavailable, the only negative reactivity is due to Doppler and moderator feedback. Consequently, the nuclear power and the transferred heat in the steam generators (SGs) increase. Since the steam flow to the turbine and the extracted power from the SGs remain constant, the SG secondary pressure and the temperatures on the primary side increase. Unless terminated by manual or automatic action, the power mismatch between primary and secondary side and the resultant coolant temperature rise could eventually result in DNB ratio and/or fuel centreline melt. In order to avoid core damage, the reactor protection system is designed to automatically

Cadmium safety rod thermal tests

International Nuclear Information System (INIS)

Thomas, J.K.; Iyer, N.C.; Peacock, H.B.

1992-01-01

Thermal testing of cadmium safety rods was conducted as part of a program to define the response of Savannah River Site (SRS) production reactor core components to a hypothetical LOCA leading to a drained reactor tank. The safety rods are present in the reactor core only during shutdown and are not used as a control mechanism during operation; thus, their response to the conditions predicted for the LOCA is only of interest to the extent that it could impact the progression of the accident. This document provides a description of this testing

Comparison of thermal behavior of different PWR fuel rod simulators for LOCA experiments

International Nuclear Information System (INIS)

Casal, V.; Malang, S.; Rust, K.

1982-10-01

For experimental investigations of a loss-of-coolant accident (LOCA) of a PWR electrical heater rods are applied as thermal fuel rod simulators. To substitute heater rods from the SEMISCALE program by INTERATOM-KfK heater rods in a current experimental program at the Instituut for Energiteknikk-(OECD-Halden), the thermodynamic behavior of different heater rods during a LOCA were compared. The results show, that SEMISCALE-heater rods can be replaced by those fabricated by INTERATOM. (orig.) [de

Control-rod, pressure and flow-induced accident and transient analysis of a direct-cycle, supercritical-pressure, light-water-cooled fast breeder reactor

International Nuclear Information System (INIS)

Kitoh, Kazuaki; Koshizuka, Seiichi; Oka, Yoshiaki

1996-01-01

The features of the direct-cycle, supercritical-pressure, light-water-cooled fast breeder reactor (SCFBR) are high thermal efficiency and simple reactor system. The safety principle is basically the same as that of an LWR since it is a water-cooled reactor. Maintaining the core flow is the basic safety requirement of the reactor, since its coolant system is the one through type. The transient behaviors at control rod, pressure and flow-induced abnormalities are analyzed and presented in this paper. The results of flow-induced transients of SCFBR were reported at ICONE-3, though pressure change was neglected. The change of fuel temperature distribution is also considered for the analysis of the rapid reactivity-induced transients such as control rod withdrawal. Total loss of flow and pump seizure are analyzed as the accidents. Loss of load, control rod withdrawal from the normal operation, loss of feedwater heating, inadvertent start of an auxiliary feedwater pump, partial loss of coolant flow and loss of external power are analyzed as the transients. The behavior of the flow-induced transients is not so much different from the analyses assuming constant pressure. Fly wheels should be equipped with the feedwater pumps to prolong the coast-down time more than 10s and to cope with the total loss of flow accident. The coolant density coefficient of the SCFBR is less than one tenth of a BWR in which the recirculation flow is used for the power control. The over pressurization transients at the loss of load is not so severe as that of a BWR. The power reaches 120%. The minimum deterioration heat flux ratio (MDHFR) and the maximum pressure are sufficiently lower than the criteria; MDHFR above 1.0 and pressure ratio below 1.10 of 27.5 MPa, maximum pressure for operation. Among the reactivity abnormalities, the control rod withdrawal transient from the normal operation is analyzed

The implication of sensitivity analysis on the safety and delayed-neutron parameters for fast breeder reactors

International Nuclear Information System (INIS)

Onega, R.J.; Florian, R.J.

1983-01-01

The delayed-neutron energy spectra for LMFBRs are not as well known as those for LWRs. These spectra are necessary for kinetics calculations which play an important role in safety and accident analyses. A sensitivity analysis was performed to study the response of the reactor power and power density to uncertainties in the delayed-neutron spectra during a rod-ejection accident. The accidents studied were central control-rod-ejections with ejection times of 2,10 and 30s. A two-energy group and two-precursor group model was formulated for the International Nuclear Fuel Cycle Evaluation (INFCE) reference design MOX-fueled LMFBR. The sensitivity analysis is based on the use of adjoints so that it is not necessary to repeatedly solve the governing (kinetics) equations to obtain the sensitivity derivatives. This is of particular importance when large systems of equations are used. The power and power-density responses were found to be most sensitive to uncertainties in the spectrum of the second delayed-neutron precursor group, resulting from the fission of 238 U, producing neutrons in the first energy group. It was found, for example, that for a rod-ejection time of 30s, and uncertainty of 7.2% in the fast components of the spectra resulted in a 24% uncertainty in the predicted power and power density. These responses were recalculated by repeatedly solving the kinetics equations. The maximum discrepancy between the recalculated and the sensitivity analysis response was only 1.6%. The results of the sensitivity analysis indicate the need for improved delayed-neutron spectral data in order to reduce the uncertainties in accident analyses. (author)

analysis of reactivity accidents in MTR for various protection system parameters and core condition

International Nuclear Information System (INIS)

Mohamed, F.M.

2011-01-01

Egypt Second Research Reactor (ETRR-2) core was modified to irradiate LEU (Low Enriched Uranium) plates in two irradiation boxes for fission 99 Mo production. The old core comprising 29 fuel elements and one Co Irradiation Device (CID) and the new core comprising 27 fuel elements, CID, and two 99 Mo production boxes. The in core irradiation has the advantage of no special cooling or irradiation loop is required. The purpose of the present work is the analysis of reactivity accidents (RIA) for ETRR-2 cores. The analysis was done to evaluate the accidents from different point of view:1- Analysis of the new core for various Reactor Protection System (RPS) parameters 2- Comparison between the two cores. 3- Analysis of the 99 Mo production boxes.PARET computer code was employed to compute various parameters. Initiating events in RIA involve various modes of reactivity insertion, namely, prompt critical condition (p=1$), accidental ejection of partial and complete CID uncontrolled withdrawal of a control rod accident, and sudden cooling of the reactor core. The time histories of reactor power, energy released, and the maximum fuel, clad and coolant temperatures of fuel elements and LEU plates were calculated for each of these accidents. The results show that the maximum clad temperatures remain well below the clad melting of both fuel and uranium plates during these accidents. It is concluded that for the new core, the RIA with scram will not result in fuel or uranium plate failure.

Control rod for FBR type reactor

International Nuclear Information System (INIS)

Nakai, Koichi.

1993-01-01

In a control rod for an LMFBR type reactor, a thermal resistor is disposed between a temperature sensitive cylinder and a cam unit support rod. A thermal expansion difference due to the temperature difference is caused between the temperature sensitive cylinder and the cam unit support rod only upon abrupt temperature change of coolants. A control rod shaft extending mechanism of downwardly depressing an absorbent portion by amplifying the thermal expansion difference by an extension link mechanism and the cam unit is provided. The thermal resistor comprises inconel 625 or like other steel of small heat conductivity. If a certain abnormality should cause to the reactor system to elevate the coolant temperature in the reactor elevates abruptly and the reactor shutdown system does not actuate, since the control rod extension shaft extends to urge the absorbent and lower the reactor core reactivity, so that leading to serious accident can be prevented surely. Further, the control rod extension shaft does not extend upon moderate temperature elevation in the usual startup and causes no unnecessary reactivity change. (N.H.)

Performance of the NRX shut-off rods

International Nuclear Information System (INIS)

Manson, R.E.

1965-08-01

A new type of shut-off rod of electromechanical design was developed by the American Machine and Foundry Company for use in the NRX reactor following the accident of 1952. The new rods were installed in May, 1956, as part of the control system conversion program which was completed in 1958. Some problems were encountered with limit switch adjustment but minor modifications in design led to much improved operation. he performance of the rods also improved as more experience was gained in the maintenance and adjustment of the various headgear components. Each headgear is now overhauled once a year on a routine basis. The present design of shut-off rod is considered to be very satisfactory. There has only been one occasion when a shut-off rod has failed to come fully down on a trip. Rods have failed to operate correctly on five other occasions but these occurred during shutdown periods or when the reactor was being shutdown manually. (author)

Assessment of WWER fuel condition in design basis accident

International Nuclear Information System (INIS)

Bibilashvili, Yu.; Sokolov, N.; Andreeva-Andrievskaya, L.; Vlasov, Yu.; Nechaeva, O.; Salatov, A.

1994-01-01

The fuel behaviour in design basis accidents is assessed by means of the verified code RAPTA-5. The code uses a set of high temperature physico-chemical properties of the fuel components as determined for commercially produced materials, fuel rod simulators and fuel rod bundles. The WWER fuel criteria available in Russia for design basis accidents do not generally differ from the similar criteria adopted for PWR's. 12 figs., 11 refs

Assessment of WWER fuel condition in design basis accident

Energy Technology Data Exchange (ETDEWEB)

Bibilashvili, Yu; Sokolov, N; Andreeva-Andrievskaya, L; Vlasov, Yu; Nechaeva, O; Salatov, A [Vsesoyuznyj Nauchno-Issledovatel` skij Inst. Neorganicheskikh Materialov, Moscow (Russian Federation)

1994-12-31

The fuel behaviour in design basis accidents is assessed by means of the verified code RAPTA-5. The code uses a set of high temperature physico-chemical properties of the fuel components as determined for commercially produced materials, fuel rod simulators and fuel rod bundles. The WWER fuel criteria available in Russia for design basis accidents do not generally differ from the similar criteria adopted for PWR`s. 12 figs., 11 refs.

The development of the fuel rod transient performance analysis code FTPAC

International Nuclear Information System (INIS)

Han Zhijie; Ji Songtao

2014-01-01

Fuel rod behavior, especially the integrity of cladding, played an important role in fuel safety research during reactor transient and hypothetical accidents conditions. In order to study fuel rod performance under transient accidents, FTPAC (Fuel Transient Performance Analysis Code) has been developed for simulating light water reactor fuel rod transient behavior when power or coolant boundary conditions are rapidly changing. It is composed of temperature, mechanical deformation, cladding oxidation and gas pressure model. The assessment was performed by comparing FTPAC code analysis result to experiments data and FRAPTRAN code calculations. Comparison shows that, the FTPAC gives reasonable agreement in temperature, deformation and gas pressure prediction. And the application of slip coefficient is more suitable for simulating the sliding between pellet and cladding when the gap is closed. (authors)

Physics analysis of the gang partial rod drive event

International Nuclear Information System (INIS)

Boman, C.; Frost, R.L.

1992-08-01

During the routine positioning of partial-length control rods in Gang 3 on the afternoon of Monday, July 27, 1992, the partial-length rods continued to drive into the reactor even after the operator released the controlling toggle switch. In response to this occurrence, the Safety Analysis and Engineering Services Group (SAEG) requested that the Applied Physics Group (APG) analyze the gang partial rod drive event. Although similar accident scenarios were considered in analysis for Chapter 15 of the Safety Analysis Report (SAR), APG and SAEG conferred and agreed that this particular type of gang partial-length rod motion event was not included in the SAR. This report details this analysis

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

International Nuclear Information System (INIS)

Otomo, Takashi; Hashimoto, Masao; Kawasaki, Satoru; Furuta, Teruo; Uetsuka, Hiroshi

1983-06-01

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/cm 2 (RT) and the heating rate was 9 0 C/s in steam with flow rate of 0.4g/cm 2 .min. The following results were obtained; (1) Maximum and burst pressures in rods were in the range 45 to 48kg/cm 2 and 41 to 45kg/cm 2 , respectively. The burst temperature of cladding were estimated to be 850 to 880 0 C. (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)

The buckling of fuel rods in transportation casks under hypothetical accident conditions

International Nuclear Information System (INIS)

Bjorkman, G.S.

2004-01-01

The buckling analysis of fuel rods during an end drop impact of a spent fuel transportation cask has traditionally been performed to demonstrate the structural integrity of the fuel rod cladding or the integrity of the fuel geometry in criticality evaluations following a cask drop event. The actual calculation of the fuel rod buckling load, however, has been the subject of some controversy, with estimates of the critical buckling load differing by as much as a factor of 5. Typically, in the buckling analysis of a fuel rod, assumptions are made regarding the percentage of fuel mass that is bonded to or participates with the cladding during the buckling process, with estimates ranging from 0 to 100%. The greater the percentage of fuel mass that is assumed to be bonded to the cladding the higher the inertia loads on the cladding, and, therefore, the lower the ''g'' value at which buckling occurs. Current published solutions do not consider displacement compatibility between the fuel and the cladding. By invoking displacement compatibility between the fuel column and the cladding column, this paper presents an exact solution for the buckling of fuel rods under inertia loading. The results show that the critical inertia load magnitude for the buckling of a fuel rod depends on the weight of the cladding and the total weight of the fuel, regardless of the percentage of fuel mass that is assumed to be attached to or participate with the cladding in the buckling process. Therefore, 100% of the fuel always participates in the buckling of a fuel rod under inertia loading

Fuel behaviour in the case of severe accidents and potential ATF designs. Fuel Behavior in Severe Accidents and Potential Accident Tolerance Fuel Designs

International Nuclear Information System (INIS)

Cheng, Bo

2013-01-01

This presentation reviews the conditions of fuel rods under severe loss of coolant conditions, approaches that may increase coping time for plant operators to recover, requirements of advanced fuel cladding to increase tolerance in accident conditions, potential candidate alloys for accident-tolerant fuel cladding and a novel design of molybdenum (Mo) -based fuel cladding. The current Zr-alloy fuel cladding will lose all its mechanical strength at 750-800 deg. C, and will react rapidly with high-pressure steam, producing significant hydrogen and exothermic heat at 700-1000 deg. C. The metallurgical properties of Zr make it unlikely that modifications of the Zr-alloy will improve the behaviour of Zr-alloys at temperatures relevant to severe accidents. The Mo-based fuel cladding is designed to (1) maintain fuel rod integrity, and reduce the release rate of hydrogen and exothermic heat in accident conditions at 1200-1500 deg. C. The EPRI research has thus far completed the design concepts, demonstration of feasibility of producing very thin wall (0.2 mm) Mo tubes. The feasibility of depositing a protective coating using various techniques has also been demonstrated. Demonstration of forming composite Mo-based cladding via mechanical reduction has been planned

Visualization test facility of nuclear fuel rod emergency cooling system

International Nuclear Information System (INIS)

Candido, Marcos Antonio; Mesquita, Amir Zacarias; Rezende, Hugo Cesar; Santos, Andre Augusto Campagnole

2013-01-01

The nuclear reactors safety is determined according to their protection against the consequences that may result from postulated accidents. The Loss of Coolant Accident (LOCA) is one the most important design basis accidents (DBA). The failure may be due to rupture of the primary loop piping. Another accident postulated is due to lack of power in the pump motors in the primary circuit. In both cases the reactor shut down automatically due to the decrease of reactivity to maintain the fissions, and to the drop of control rods. In the event of an accident it is necessary to maintain the coolant flow to remove the fuel elements residual heat, which remains after shut down. This heat is a significant amount of the maximum thermal power generated in normal operation (about 7%). Recently this event has been quite prominent in the press due to the reactor accident in Fukushima nuclear power station. This paper presents the experimental facility under rebuilding at the Thermal Hydraulic Laboratory of the Nuclear Technology Development Center (CDTN) that has the objective of monitoring and visualization of the process of emergency cooling of a nuclear fuel rod simulator, heated by Joule effect. The system will help the comprehension of the heat transfer process during reflooding after a loss of coolant accident in the fuel of light water reactor core. (author)

Enhancement of weld failure and tube ejection model in PENTAP program

International Nuclear Information System (INIS)

Jung, Jaehoon; An, Sang Mo; Ha, Kwang Soon; Kim, Hwan Yeol

2014-01-01

The reactor vessel pressure, the debris mass, the debris temperature, and the component of material can have an effect on the penetration tube failure modes. Furthermore, these parameters are interrelated. There are some representative severe accident codes such as MELCOR, MAAP, and PENTAP program. MELCOR decides on a penetration tube failure by its failure temperature such as 1273K simply. MAAP considers all penetration failure modes and has the most advanced model for a penetration tube failure model. However, the validation work against the experimental data is very limited. PENTAP program which evaluates the possible penetration tube failure modes such as creep failure, weld failure, tube ejection, and a long term tube failure under given accident condition was developed by KAERI. The experiment for the tube ejection is being performed by KAERI. The temperature distribution and the ablation rate of both weld and lower vessel wall can be obtained through the experiment. This paper includes the updated calculation steps for the weld failure and the tube ejection modes of the PENTAP program to apply the experimental results. PENTAP program can evaluate the possible penetration tube failure modes. It still requires a large amount of efforts to increase the prediction of failure modes. Some calculation steps are necessary for applying the experimental and the numerical data in the PENTAP program. In this study, new calculation steps are added to PENTAP program to enhance the weld failure and tube ejection models using KAERI's experimental data which are the ablation rate and temperature distribution of weld and lower vessel wall

Risk evaluation of accident management strategies

International Nuclear Information System (INIS)

Dingman, S.; Camp, A.

1992-01-01

The use of Probabilistic Risk Assessment (PRA) methods to evaluate accident management strategies in nuclear power plants discussed in this paper. The PRA framework allows an integrated evaluation to be performed to give the full implications of a particular strategy. The methodology is demonstrated for a particular accident management strategy, intentional depressurization of the reactor coolant system to avoid containment pressurization during the ejection of molten debris at vessel breach

Rod bundle burnout data and correlation comparisons

International Nuclear Information System (INIS)

Yoder, G.L.; Morris, D.G.; Mullins, C.B.

1985-01-01

Rod bundle burnout data from 30 steady-state and 3 transient tests were obtained from experiments performed in the Thermal Hydraulic Test Facility at the Oak Ridge National Laboratory. The tests covered a parameter range relevant to intact core reactor accidents ranging from large break to small break loss-ofcoolant conditions. Instrumentation within the 64-rod test section indicated that burnout occurred over an axial range within the bundle. The distance from the point where the first dry rod was detected to the point where all rods were dry was up to 60 cm in some of the tests. The burnout data should prove useful in developing new correlations for use in reactor thermalhydraulic codes. Evaluation of several existing critical heat flux correlations using the data show that three correlations, the Barnett, Bowring, and Katto correlations, perform similarly and correlate the data better than the Biasi correlation

Fuel models and results from the TRAC-PF1/MIMAS TMI-2 accident calculation

International Nuclear Information System (INIS)

Schwegler, E.C.; Maudlin, P.J.

1983-01-01

A brief description of several fuel models used in the TRAC-PF1/MIMAS analysis of the TMI-2 accident is presented, and some of the significant fuel-rod behavior results from this analysis are given. Peak fuel-rod temperatures, oxidation heat production, and embrittlement and failure behavior calculated for the TMI-2 accident are discussed. Other aspects of fuel behavior, such as cladding ballooning and fuel-cladding eutectic formation, were found not to significantly affect the accident progression

Historical survey of the qualifying process of Furnas calculus methodology in the areas of rods, neutronics, thermohydraulic accidents and transients

International Nuclear Information System (INIS)

Conti, C.F.S.; Silva Galetti, M.R. da.

1990-02-01

As Furnas intends to assume in the future the responsibility of performing Safety Analyses associated to Reload and Operation questions to Angra 1, it was figured out the necessity of qualifying its methodology by CNEN. The Methodology Qualification Process is based on guidelines proposed by CNEN at NT-DR-N o 02/87, where it was divided in four steps. This Technical Note aims to present the follow up of FURNAS Methodology Qualification Process and to bring it up to date in the areas of Core Physics (Neutronics), Core Thermal-Hydraulics, Fuel Rod Behaviour, Transient and Large Break Loss of Coolant Accident Analyses (LBLOCA). (author)

Analysis and sensitivity studies with CORETRAN and RETRAN-3D of the NEACRP PWR rod ejection benchmark

Energy Technology Data Exchange (ETDEWEB)

Ferroukhi, H.; Coddington, P. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

2001-07-01

The OECD/NEA PWR rod ejection benchmark has been analysed using the 3-D nodal spatial-kinetic codes CORETRAN and RETRAN-3D. The following results were obtained. A) The agreement in 3-D solution between CORETRAN and RETRAN-3D was found to be very good both during steady-state and transient conditions. In particular at HZP (hot zero power), an excellent agreement in the initial steady-state 3-D power distribution and with regard to the core power excursion during the super-prompt critical phase of the transient (i.e. when the negative reactivity feedback is still very weak) was found. This illustrates the consistency in the neutronic solution between both codes. B) At both HZP and FP (full power) conditions, the CORETRAN and RETRAN-3D results lie well within the range of the previous benchmark solutions. In particular at HZP, both codes predict a power excursion and an increase in maximum pellet temperature that are among the closest results to those obtained with the benchmark reference solution. It must here be emphasised that these analyses are by no means a validation of the codes. However, the good agreement of both CORETRAN and RETRAN-3D with other 3-D solutions provides confidence in the ability of these codes to analyse LWR (light water reactor) core transients. In addition, it was found appropriate to perform, for this well-defined international benchmark problem, some sensitivity studies in order to assess the impact of modelling options on the CORETRAN and RETRAN-3D results. (authors)

Analysis and sensitivity studies with CORETRAN and RETRAN-3D of the NEACRP PWR rod ejection benchmark

International Nuclear Information System (INIS)

Ferroukhi, H.; Coddington, P.

2001-01-01

The OECD/NEA PWR rod ejection benchmark has been analysed using the 3-D nodal spatial-kinetic codes CORETRAN and RETRAN-3D. The following results were obtained. A) The agreement in 3-D solution between CORETRAN and RETRAN-3D was found to be very good both during steady-state and transient conditions. In particular at HZP (hot zero power), an excellent agreement in the initial steady-state 3-D power distribution and with regard to the core power excursion during the super-prompt critical phase of the transient (i.e. when the negative reactivity feedback is still very weak) was found. This illustrates the consistency in the neutronic solution between both codes. B) At both HZP and FP (full power) conditions, the CORETRAN and RETRAN-3D results lie well within the range of the previous benchmark solutions. In particular at HZP, both codes predict a power excursion and an increase in maximum pellet temperature that are among the closest results to those obtained with the benchmark reference solution. It must here be emphasised that these analyses are by no means a validation of the codes. However, the good agreement of both CORETRAN and RETRAN-3D with other 3-D solutions provides confidence in the ability of these codes to analyse LWR (light water reactor) core transients. In addition, it was found appropriate to perform, for this well-defined international benchmark problem, some sensitivity studies in order to assess the impact of modelling options on the CORETRAN and RETRAN-3D results. (authors)

Fuel rod behavior of a PWR during load following

International Nuclear Information System (INIS)

Perrotta, J.A.; Andrade, G.G. de

1982-01-01

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.) [pt

Computer code for the analysis of destructive pressure generation process during a fuel failure accident, PULSE-2

International Nuclear Information System (INIS)

Fujishiro, Toshio

1978-03-01

The computer code PULSE-2 has been developed for the analysis of pressure pulse generation process when hot fuel particles come into contact with the coolant in a fuel rod failure accident. In the program, it is assumed that hot fuel fragments mix with the coolant instantly and homogeneously in the failure region. Then, the rapid vaporization of the coolant and transient pressure rise in failure region, and the movement of ejected coolant slugs are calculated. The effect of a fuel-particle size distribution is taken into consideration. Heat conduction in the fuel particles and heat transfer at fuel-coolant interface are calculated. Temperature, pressure and void fraction in the mixed region are calculated from the average enthalpy. With physical property subroutines for liquid sodium and water, the model is usable for both LMFBR and LWR conditions. (auth.)

Pilot ejection, parachute, and helicopter crash injuries.

Science.gov (United States)

McBratney, Colleen M; Rush, Stephen; Kharod, Chetan U

2014-01-01

USAF Pararescuemen (PJs) respond to downed aircrew as a fundamental mission for personnel recovery (PR), one of the Air Force's core functions. In addition to responding to these in Military settings, the PJs from the 212 Rescue Squadron routinely respond to small plane crashes in remote regions of Alaska. While there is a paucity of information on the latter, there have been articles detailing injuries sustained from helicopter crashes and while ejecting or parachuting from fixed wing aircraft. The following represents a new chapter added to the Pararescue Medical Operations Handbook, Sixth Edition (2014, editors Matt Wolf, MD, and Stephen Rush, MD, in press). It was designed to be a quick reference for PJs and their Special Operations flight surgeons to help with understanding of mechanism of injury with regard to pilot ejection, parachute, and helicopter accident injuries. It outlines the nature of the injuries sustained in such mishaps and provides an epidemiologic framework from which to approach the problem. 2014.

Ameliorative design for CARR safety rod drive mechanism

International Nuclear Information System (INIS)

Zhu Xuewei; Luo Zhong; Zhen Jianxiao; Wang Yulin

2014-01-01

The problem of safety rod accident dropped during C commissioning phase for China Advanced Research Reactor (CARR) was analyzed, and the reason was that the solenoid valve in safety rod drive mechanism (SRDM) driven loop was breakdown because of long-playing work. To solve this safe hidden trouble, SRDM was redesigned, and a new type of 'hydro lifting-hydro and electromagnetic holding' SRDM was presented, using Ansoft Maxwell to make a finite element analysis on new SRDM, working out electromagnetic field distribution and electromagnetic force of new SRDM. The results show that the value of electromagnetic force produced by electromagnetic force holding unit reaches 2.12 times about the weight of safety rod drive line, and it has some margins. (authors)

Behavior of a corium jet in high pressure melt ejection from a reactor pressure vessel

International Nuclear Information System (INIS)

Frid, W.

1988-04-01

Discharge of the molten core debris from a pressurized reactor vessel has been recognized as an important accident scenario for pressurized water reactors. Recent high-pressure melt streaming experiments conducted at Sandia National Laboratories, designed to study cavity and containment events related to melt ejection, have resulted in two important observations: (1) Expansion and breakup of the ejected molten jet. (2) Significant aerosol generation during the ejection process. The expansion and breakup of the jet in the experiments are attributed to rapid evolution of the pressurizing gas (nitrogen or hydrogen) dissolved in the melt. It has been concluded that aerosol particles may be formed by condensation of melt vapor and mechanical breakup of the melt and generation. It was also shown that the above stated phenomena are likely to occur in reactor accidents. This report provides results from analytical and experimental investigations on the behavior of a gas supersaturated molten jet expelled from a pressurized vessel. Aero-hydrodynamic stability of liquid jets in gas, stream degassing of molten metals, and gas bubble nucleation in molten metals are relevant problems that are addressed in this work

Accident analysis of RB reactor dependent on the lattice pitch; Akcidentalna analiza reaktora ''RB'' pri promeni koraka resetke

Energy Technology Data Exchange (ETDEWEB)

Lolic, B; Lazarevic, B [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

1963-07-01

This analysis was concerned with reactor core with 52-56 fuel rods, lattice pitch being, 8, 10, 12, 16, 18, and 20 cm. Measured values of excess reactivity above critical level of 3.85 cm, total anti reactivity of regulating rod, reactivity changes caused by pumping heavy water and reactivity variations due to movement of control rod were used. Three types of accidents were analyzed: movement of regulating rod to the position of zero reactivity worths, increase of heavy water level at rate of 2.5 cm/min, combination of two previous accidents.

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

International Nuclear Information System (INIS)

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

1979-01-01

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

Annular burnout data from rod-bundle experiments

International Nuclear Information System (INIS)

Yoder, G.L.; Morris, D.G.; Mullins, C.B.

1983-01-01

Burnout data for annular flow in a rod bundle are presented for both transient and steady-state conditions. Tests were performed at the Oak Ridge National Laboratory in the Thermal Hydraulic Test Facility (THTF), a pressurized-water loop containing an electrically heated 64-rod bundle. The bundle configuration is typical of later generation pressurized-water reactors with 17 x 17 fuel arrays. Both axial and radial power profiles are flat. All experiments were carried out in upflow with subcooled inlet conditions, insuring accurate flow measurement. Conditions within the bundle were typical of those which could be encountered during a nuclear reactor loss-of-coolant accident

Fuel-rod response during the large-break LOCA Test LOC-6

International Nuclear Information System (INIS)

Vinjamuri, K.; Cook, B.A.; Hobbins, R.R.

1981-01-01

The large break Loss of Coolant Accident (LOCA) Test LOC-6 was conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory by EG and G Idaho, Inc. The objectives of the PBF LOCA tests are to obtain in-pile cladding ballooning data under blowdown and reflood conditions and assess how well out-of-pile ballooning data represent in-pile fuel rod behavior. The primary objective of the LOC-6 test was to determine the effects of internal rod pressures and prior irradiation on the deformation behavior of fuel rods that reached cladding temperatures high in the alpha phase of zircaloy. Test LOC-6 was conducted with four rods of PWR 15 x 15 design with the exception of fuel stack length (89 cm) and enrichment (12.5 W% 235 U). Each rod was surrounded by an individual flow shroud

Evaluation of fuel rod damage in LWR under accident conditions using SSYST

International Nuclear Information System (INIS)

Meyder, R.

1982-01-01

After a short outline of the recent SSYST-development, the creep rupture model NORA 2 is presented. The effect of temperature and oxygen on Zircaloy 4 creep behaviour is shown. Examples on the effect of azimuthal varying gap width and wall thickness are given. Remarks on the extension of a single rod analysis on a bundle and the stepwise application of SSYST for investigation of fuel rod failure conclude the paper. (orig.) [de

Analysis of reactivity insertion accidents in PWR reactors

International Nuclear Information System (INIS)

Camargo, C.T.M.

1978-06-01

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 UO 2 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) [pt

Degradation in steam of 60 cm-long B{sub 4}C control rods

Energy Technology Data Exchange (ETDEWEB)

Dominguez, C., E-mail: christina.dominguez@irsn.fr; Drouan, D.

2014-08-01

In the framework of nuclear reactor core meltdown accident studies, the degradation of boron carbide control rod segments exposed to argon/steam atmospheres was investigated up to about 2000 °C in IRSN laboratories. The sequence of the phenomena involved in the degradation has been found to take place as expected. Nevertheless, the ZrO{sub 2} oxide layer formed on the outer surface of the guide tube was very protective, significantly delaying and limiting the guide tube failure and therefore the boron carbide pellet oxidation. Contrary to what was expected, the presence of the control rod decreases the hydrogen release instead of increasing it by additional oxidation of boron compounds. Boron contents up to 20 wt.% were measured in metallic mixtures formed during degradation. It was observed that these metallic melts are able to attack the surrounding fuel rods, which could have consequences on fuel degradation and fission product release kinetics during severe accidents.

Analytical functions used for description of the plastic deformation process in Zirconium alloys WWER type fuel rod cladding under designed accident conditions

International Nuclear Information System (INIS)

Fedotov, A.

2003-01-01

The aim of this work was to improve the RAPTA-5 code as applied to the analysis of the thermomechanical behavior of the fuel rod cladding under designed accident conditions. The irreversible process thermodynamics methods were proposed to be used for the description of the plastic deformation process in zirconium alloys under accident conditions. Functions, which describe yielding stress dependence on plastic strain, strain rate and temperature may be successfully used in calculations. On the basis of the experiments made and the existent experimental data the dependence of yielding stress on plastic strain, strain rate, temperature and heating rate for E110 alloy was determined. In future the following research work shall be made: research of dynamic strain ageing in E635 alloy under different strain rates; research of strain rate influence on plastic strain in E635 alloy under test temperature higher than 873 K; research of deformation strengthening of E635 alloy under high temperatures; research of heating rate influence n phase transformation in E110 and E635 alloys

Development of in-vessel type control rod drive mechanism for a innovative small reactor (Contract research)

Energy Technology Data Exchange (ETDEWEB)

Yoritsune, Tsutomu; Ishida, Toshihisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2003-03-01

Although the control rod drive mechanism of an existing large scale light water reactor is generally installed outside the reactor vessel, an in-vessel type control rod drive mechanism (INV-CRDM) is installed inside the reactor vessel. The INV-CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. Therefore, INV-CRDM is an important technology adopted in an innovative small reactor. Japan Atomic Energy Research Institute (JAERI) has developed this type of CRDM driven by an electric motor, which can work under high temperature and high pressure water for the advanced marine reactor. On the basis of this research result, a driving motor coil and a bearing were developed to be used under the high temperature steam, severe condition for an innovative small reactor. About the driving motor, we manufactured the driving motor available for high temperature steam and carried out performance test under room temperature atmosphere to confirm the electric characteristic and coolability of the driving coil. With these test results and the past test results under high temperature water, we analyzed and evaluated the electric performance and coolability of the driving coil under high temperature steam. Concerning bearing, we manufactured the test pieces using some candidate material for material characteristic test and carried out the rolling wear test under high temperature steam to select the material. Consequently, we confirmed that performance of the driving coil for the advanced type driving motor, is enough to be used under high temperature steam. And, we evaluated the performance of the bearing and selected the material of the bearing, which can be used under high temperature steam. From these results, we have obtained the prospect that the INV-CRDM can be used for an innovative small reactor under steam atmosphere could be developed. (author)

Substitute safety rods: Physics of operation and irradiation

International Nuclear Information System (INIS)

Baumann, N.P.

1991-01-01

Under certain assumed accidents, an SRS reactor may lose most of its bulk moderator while maintaining flow to fuel assemblies. If this occurs immediately after operation at power, components normally dependent on convective heat transfer to the moderator will heat up with the possibility of melting that component. One component at risk is the currently used cadmium safety rod. A substitute safety rod consisting solely of sintered B 4 C and stainless steel has been designed which is capable of withstanding much higher temperatures. This memorandum provides the physics basis for the adequacy of the rod for reactor shutdown and provides a set of criteria for acceptance in the NTG tests. This memorandum provides physics data for other aspects of operation. These include: Heat production and helium production, along with related phenomena, resulting from inadvertent irradiation at power. Gamma heat input under drained tank conditions. An equivalent rod design suitable for charge design and safety analyses. Degradation under normal operation. Thermal flux ripple in adjacent fuel due to axial striping of alternate B 4 C and steel pellets. Possible effect on safety analyses. Safety rod withdrawal during reactor startup

Investigation of spatial coupling aspects for coupled code application in PWR safety analysis

International Nuclear Information System (INIS)

Todorova, N.K.; Ivanov, K.N.

2003-01-01

The simulation of nuclear power plant accident conditions requires three-dimensional (3-D) modeling of the reactor core to ensure a realistic description of physical phenomena. This paper describes a part of the research activities carried out on the sensitivity of coupled neutronics/thermal-hydraulic system code's results to the spatial mesh overlays used for modeling pressurized water reactor (PWR) cores for analysis of different transients. The coupled TRAC-PF1/NEM was used to model PWR rod ejection accident (REA). Modeling schemes for pressurized water reactor are described in detail, followed by a comparative analysis of both steady state and transient calculations. By using different TRAC-PF1/NEM vessel modeling options it was demonstrated that the geometric refinement plays a great role in determining the local parameters and control rod worth in the case of spatially asymmetric transients. The capability of TRAC-PF1/NEM to introduce local refinement of heat structure models was explored while preserving the original coarse-mesh structure of the hydraulic model. The obtained results indicated that the thermal-hydraulic feedback phenomenon is non-linear and cannot be separated even in rod ejection accident analysis, where the Doppler feedback plays a dominant role. While the impact of neutronics mesh refinement is well known, this research found that the local predictions, as well as the global predictions are also very sensitive to the thermal-hydraulic refinement

Study of corium radial spreading between fuel rods in a PWR core

International Nuclear Information System (INIS)

Roche, S.; Gatt, J.M.

1996-01-01

In the framework of severe accident studies for PWR like Three Mile Island Unit 2 (TMI-2), the reactor core essentially constituted of fuel rods begins to heat and then to melt. During the early degradation phase, a melt (essentially UO2 and ZrO2) that constitutes the corium flows first along the rods, and after a blockage formation, may radially propagate towards the core periphery. A simplified model has been elaborated to study the corium freezing phenomena during its crossflow between the fuel rods. The corium spreads on an horizontal support made, of either a corium crust, or a grid assembly. The model solves numerically the interface energy balance equation at the solid-liquid corium interface and the monodimensional heat balance equation in transient process with convective terms and heat source (residual power). ''Zukauskas'' correlations are used to calculate heat transfer coefficients. The model can be integrated in severe accident codes like ICARE II (IPSN) describing the in-vessel degradation scenarios. (author). 5 refs, 10 figs

Axial distribution of deformation in the cladding of pressurized water reactor fuel rods in a loss-of-coolant accident

International Nuclear Information System (INIS)

Rose, K.M.; Mann, C.A.; Hindle, E.D.

1979-01-01

In the event of a loss-of-coolant accident in a pressurized water reactor, the cladding of the fuel rods would undergo a temperature excursion while being subject to tensile hoop stress. The deformation behavior of 470-mm lengths of Zircaloy-4 fuel cladding has been studied experimentally; under a range of stress levels in the high-alpha range of zirconium (600 to 850 0 C), diametral strains of up to 70% were observed over the greater part of their length. A negative-feedback mechanism is suggested, based on the reduction of secondary creep rate following cooling by enhanced heat loss at swelling areas. An approximate analysis based on this mechanism was found to be in reasonable agreement with the experimental results. A computer modeling code is being developed to predict cladding deformation under realistic conditions

Axial distribution of deformation in the cladding of pressurized water reactor fuel rods in a loss-of-coolant accident

Energy Technology Data Exchange (ETDEWEB)

Rose, K.M.; Mann, C.A.; Hindle, E.D.

1979-12-01

In the event of a loss-of-coolant accident in a pressurized water reactor, the cladding of the fuel rods would undergo a temperature excursion while being subject to tensile hoop stress. The deformation behavior of 470-mm lengths of Zircaloy-4 fuel cladding has been studied experimentally; under a range of stress levels in the high-alpha range of zirconium (600 to 850/sup 0/C), diametral strains of up to 70% were observed over the greater part of their length. A negative-feedback mechanism is suggested, based on the reduction of secondary creep rate following cooling by enhanced heat loss at swelling areas. An approximate analysis based on this mechanism was found to be in reasonable agreement with the experimental results. A computer modeling code is being developed to predict cladding deformation under realistic conditions.

Multi-rod burst behavior under a loss-of-coolant accident condition, (1)

International Nuclear Information System (INIS)

Hashimoto, Masao; Otomo, Takashi; Furuta, Teruo; Kawasaki, Satoru; Uetsuka, Hiroshi

1980-12-01

Multi-rod burst tests have been planned since 1977 to estimate quantitative channel restriction during a LOCA transient in LWRs. For this purpose, many bundle tests have been making to burst in a steam in varying a few parameters which influence the degree of channel restriction. The purpose of this report is to provide a background document for final reports to be published in the future. This report includes the results of No. 7805 bundle test, namely temperature, internal pressure, burst behavior of rods and channel restriction of the bundle. (author)

PBF/LOFT Lead Rod Test Program experiment predictions document

International Nuclear Information System (INIS)

Varacalle, D.J.; Cox, W.R.; Niebruegge, D.A.; Seiber, S.J.; Brake, T.E.; Driskell, W.E.; Nigg, D.W.; Tolman, E.L.

1978-12-01

The PBF/LOFT Lead Rod (LLR) Test Program is being conducted to provide experimental information on the behavior of nuclear fuel under normal and accident conditions in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. The PBF/LLR tests are designed to simulate the test conditions for the LOFT Power Ascension Tests L2-3 through L2-5. The test program has been designed to provide a parametric evaluation of the LOFT fuel (center and peripheral modules) over a wide range of power. This report presents the experiment predictions for the three four-rod LOCA tests

Fluid Damping Variation of a Slender Rod in Axial Flow Field

Energy Technology Data Exchange (ETDEWEB)

Park, Nam-Gyu; Yoo, Jong-Sung; Jung, Yil-Sup [KEPCO Nuclear Fuel Co., Daejeon (Korea, Republic of)

2016-10-15

This study proposed an analytic damping model considering the axial flow condition. In addition, the specific damping values with respect to the flow speeds are calculated. The flow induced damping is beneficial to fuel integrity in that impact energy due to severe accidents such as earthquake dissipates rapidly. A nuclear fuel bundle is composed of many slender fuel rods which contain fission material. The slender rod is typical structure in the fuel, therefore fluid damping estimation on the rod should be an important clue leading to fuel bundle damping identification. Severe accidents could cause fuel assembly vibration in the core, but large motion could be damped out rapidly when a strong damping mechanism is involved. This paper suggested a mathematical model of the slender structure. The physical meaning of the model is described, and the simulation results with the model are also provided. Actual damping due to the fluid is nonlinear, therefore further works are required to explain the detail behavior with the nonlinearity. The model validation test is on-going in KEPCO Nuclear Fuel, but it is believed that performance of the model is well correlated to the published work.

Physics calculations for the RIA 1-3 irradiated rod test

International Nuclear Information System (INIS)

Young, T.E.

1981-06-01

The RIA 1-3 test would employ a square array of four pre-irradiated BWR rods to provide information on fuel failure modes and consequences of postulated Reactivity Initiated Accidents in power reactors. Calculations were done to: (1) predict R-O power distributions in the test rods for thermal-hydraulic and fuel-failure analysis; and (2) predict the steady-state and transient ratios of test fuel energy deposition to core energy deposition (Figures of Merit). Fission distributions for the test were computed with the RAFFL Monte Carlo code using an external neutron current source from a complete-reactor radial calculation with the SCAMP S/sub n/ code. Energies per fission for the rods were computed using the SINBAD buildup and depletion code, the GAMSOR gamma ray source code, and the QAD-BSA point-kernel shielding code. The calculated rod average-to-test average energy deposition ratios are 0.99, 0.99, and 0.97 for the rods irradiated to approximately 12 CWd/tu, and 1.04 for the rod irradiated to 4.8 GWd/tu. The maximum deviation of the power density of 1/12-rod azimuthal segments from the rod average is 4%. For an estimated control rod position of 0.591 m withdrawn the predicted radial average energy deposition at the axial peak in an average test rod is 1.71 (kW/m)/MW during preconditioning, and 1.84 (kJ/kg UO 2 ) MW.S during the burst. 16 figures, 7 tables

Examination of cadmium safety rod thermal test specimens and failure mechanism evaluation

International Nuclear Information System (INIS)

Thomas, J.K.; Peacock, H.B.; Iyer, N.C.

1992-01-01

The reactor safety rods may be subjected to high temperatures due to gamma heating after the core coolant level has dropped during the ECS phase of a hypothetical LOCA event. Accordingly, an experimental cadmium safety rod testing subtask was established as part of a task to address the response of reactor core components to this accident. Companion reports describe the experiments and a structural evaluation (finite element analysis) of the safety rod. This report deals primarily with the examination of the test specimens, evaluation of possible failure mechanisms, and confirmatory separate effects experiments. It is concluded that the failures observed in the cadmium safety rod thermal tests which occurred at low temperature (T 800 degrees C) with fast thermal ramp rates are concluded to be mechanical in nature without significant environmental degradation. Based on these tests, tasks were initiated to design and manufacture B 4 C safety rods to replace the cadmium safety rods. The B 4 C safety rods have been manufactured at this time and it is currently planned to charge them to the reactor in the near future. 60 refs

Reactor core and control rod assembly in FBR type reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi.

1993-01-01

Fuel assemblies and control rod assemblies are attached respectively to reactor core support plates each in a cantilever fashion. Intermediate spacer pads are disposed to the lateral side of a wrapper tube just above the fuel rod region. Intermediate space pads are disposed to the lateral side of a control rod guide tube just above a fuel rod region. The thickness of the intermediate spacer pad for the control rod assembly is made smaller than the thickness of the intermediate spacer pad for the fuel assembly. This can prevent contact between intermediate spacer pads of the control guide tube and the fuel assembly even if the temperature of coolants is elevated to thermally expand the intermediate spacer pad, by which the radial displacement amount of the reactor core region along the direction of the height of the control guide tube is reduced substantially to zero. Accordingly, contribution of the control rod assembly to the radial expansion reactivity can be reduced to zero or negative level, by which the effect of the negative radial expansion reactivity of the reactor is increased to improve the safety upon thermal transient stage, for example, loss of coolant flow rate accident. (I.N.)

Assessment of stainless steel 348 fuel rod performance against literature available data using TRANSURANUS code

Directory of Open Access Journals (Sweden)

Giovedi Claudia

2016-01-01

Full Text Available Early pressurized water reactors were originally designed to operate using stainless steel as cladding material, but during their lifetime this material was replaced by zirconium-based alloys. However, after the Fukushima Daiichi accident, the problems related to the zirconium-based alloys due to the hydrogen production and explosion under severe accident brought the importance to assess different materials. In this sense, initiatives as ATF (Accident Tolerant Fuel program are considering different material as fuel cladding and, one candidate is iron-based alloy. In order to assess the fuel performance of fuel rods manufactured using iron-based alloy as cladding material, it was necessary to select a specific stainless steel (type 348 and modify properly conventional fuel performance codes developed in the last decades. Then, 348 stainless steel mechanical and physics properties were introduced in the TRANSURANUS code. The aim of this paper is to present the obtained results concerning the verification of the modified TRANSURANUS code version against data collected from the open literature, related to reactors which operated using stainless steel as cladding. Considering that some data were not available, some assumptions had to be made. Important differences related to the conventional fuel rods were taken into account. Obtained results regarding the cladding behavior are in agreement with available information. This constitutes an evidence of the modified TRANSURANUS code capabilities to perform fuel rod investigation of fuel rods manufactured using 348 stainless steel as cladding material.

Simulation of LOF accidents with directly electrical heated UO2 pins

International Nuclear Information System (INIS)

Alexas, A.

1976-01-01

The behavior of directly electrical heated UO 2 pins has been investigated under loss of coolant conditions. Two types of hypothetical accidents have been simulated, first, a LOF accident without power excursion (LOF accident) and second, a LOF accident with subsequent power excursion (LOF-TOP accident). A high-speed film shows the sequence of events for two characteristic experiments. In consequence of the high-speed film analysis as well as the metallographical evaluation statements are given in respect to the cladding meltdown process, the fuel melt fraction and the energy input from the beginning of a power transient to the beginning of the molten fuel ejections

Three-dimensional space-time kinetic analysis with CORETRAN and RETRAN-3D of the NEACRP PWR rod ejection benchmark

Energy Technology Data Exchange (ETDEWEB)

Ferroukhi, H.; Coddington, P

2001-03-01

One of the activities within the STARS project, in the Laboratory for Reactor Physics and System Behaviour; is the development of a coupling methodology between the three-dimensional, space-time kinetics codes CORETRAN and RETRAN-3D in order to perform core and plant transient analyses of the Swiss LWRs. The CORETRAN code is a 3-D full-core simulator, intended to be used for core-related analyses, while RETRAN-3D is the three-dimensional kinetics version of the plant system code RETRAN, and can therefore be used for best-estimate analyses of a wide range of transients in both PWRs and BWRs. Because the neutronics solver in both codes is based on the same kinetics model, one important advantage is that the codes can be coupled so that the initial conditions for a RETRAN-3D plant analysis are generated by a detailed-core, steady-state calculation using CORETRAN. As a first step towards using CORETRAN and RETRAN-3D for kinetic applications, the NEACRP PWR rod ejection benchmark has been analyzed with both codes, and is presented in this paper. The first objective is to verify the consistency between the static and kinetic solutions of the two codes, and so gain confidence in the coupling methodology. The second objective is to assess the CORETRAN and RETRAN-3D solutions for a well-defined RIA transient, comparing with previously published results. In parallel, several sensitivity studies have been performed in an attempt to identify models and calculational options important for a correct analysis of an RIA event in a LWR using these two codes. (author)

Department of Reactor Technology: annual progress report 1 January - 31 December 1976

International Nuclear Information System (INIS)

1977-06-01

The work of the Department of Reactor Technology within the following fields is described: reactor engineering, structural reliability, system reliability, radiation fiels in nuclear power plants, reactor physics, fuel management, fission product decay analysis, steady-state thermo-hydraulics, reactor accident analysis for LOCA and ECC, containment analysis, experimental heat transfer, reactor core dynamics and power plant simulators, control rod ejection accident analysis, economic studies for power plants, experimental activation measurements and neutron radiography at the DR 1 reactor. (author)

Models of multi-rod code FRETA-B for transient fuel behavior analysis

International Nuclear Information System (INIS)

Uchida, Masaaki; Otsubo, Naoaki.

1984-11-01

This paper is a final report of the development of FRETA-B code, which analyzes the LWR fuel behavior during accidents, particularly the Loss-of-Coolant Accident (LOCA). The very high temperature induced by a LOCA causes oxidation of the cladding by steam and, as a combined effect with low external pressure, extensive swelling of the cladding. The latter may reach a level that the rods block the coolant channel. To analyze these phenomena, single-rod model is insufficient; FRETA-B has a capability to handle multiple fuel rods in a bundle simultaneously, including the interaction between them. In the development work, therefore, efforts were made for avoiding the excessive increase of calculation time and core memory requirement. Because of the strong dependency of the in-LOCA fuel behavior on the coolant state, FRETA-B has emphasis on heat transfer to the coolant as well as the cladding deformation. In the final version, a capability was added to analyze the fuel behavior under reflooding using empirical models. The present report describes the basic models of FRETA-B, and also gives its input manual in the appendix. (author)

Reactivity Accidents in CAREM-25 Core with and Without Safety Systems Actuation

International Nuclear Information System (INIS)

Gimenez, Marcelo; Vertullo, Alicia; Schlamp, Miguel

2000-01-01

A reactivity accident in CAREM core can be provoked by different initiating events, a cold water injection in pressure vessel, a secondary side steam line breakage and a failure in the absorbing rods drive system.The present work analyses inadverted control rod withdraws transients.Maximum worth control rod (2.5 $) at normal velocity (1 cm/s) is adopted for the simulations (Reactivity ramp of 0.018 $/s).Different scenarios considering actuation of first shutdown system (FSS), second shutdown system (SSS) and selflimiting conditions were modeled.Results of the accident with actuation of FSS show that safety margins are well above critical values (DNBR and CPR).In the cases with failure of the FSS and success of SSS or selflimited, safety margins are below critical values, however, the SSS provides a reduction of elapsed time under advised margins

Severe accident recriticality analyses (SARA)

DEFF Research Database (Denmark)

Frid, W.; Højerup, C.F.; Lindholm, I.

2001-01-01

with all three codes. The core initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality-both super-prompt power bursts and quasi steady-state power......Recriticality in a BWR during reflooding of an overheated partly degraded core, i.e. with relocated control rods, has been studied for a total loss of electric power accident scenario. In order to assess the impact of recriticality on reactor safety, including accident management strategies......, which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal g(-1), was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate of 2000 kg s(-1). In most cases, however, the predicted energy deposition was smaller, below...

CFD analysis of blockage length on a partially blocked fuel rod

International Nuclear Information System (INIS)

Scuro, Nikolas Lymberis; Andrade, Delvonei Alves de; Angelo, Gabriel; Angelo, Edvaldo

2017-01-01

In LOCA accidents, fuel rods may balloon by the increasing of pressure difference between fuel rod and core vessel. With the balloon effect, the swelling can partially block the flow channel, affecting the coolability during reflood phase. In order to analyze the influence of blockage length after LOCA events, many numerical simulations using Ansys-CFX code have been done in steady state condition, characterizing the final phase of reflood. Peaks of temperature are observed in the middle of the fuel rod, followed by a temperature drop. This effect is justified by the increasing of heat transfer coefficient, originated from the high turbulence effects. Therefore, this paper considers a radial blockage of 90%, varying just the blockage length. This study observed that, for the same boundary conditions, the longer the blockage length originated after LOCA events, the higher are the central temperatures in the fuel rod. (author)

CFD analysis of blockage length on a partially blocked fuel rod

Energy Technology Data Exchange (ETDEWEB)

Scuro, Nikolas Lymberis; Andrade, Delvonei Alves de [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear; Angelo, Gabriel [Centro Universitário FEI (UNIFEI), São Paulo, SP (Brazil). Dept. de Engenharia Mecânica; Angelo, Edvaldo, E-mail: nikolas.scuro@gmail.com, E-mail: delvonei@ipen.br, E-mail: gangelo@fei.edu.br, E-mail: eangelo@mackenzie.br [Universidade Presbiteriana Mackenzie, São Paulo, SP (Brazil). Escola da Engenharia. Grupo de Simulação Numérica

2017-07-01

In LOCA accidents, fuel rods may balloon by the increasing of pressure difference between fuel rod and core vessel. With the balloon effect, the swelling can partially block the flow channel, affecting the coolability during reflood phase. In order to analyze the influence of blockage length after LOCA events, many numerical simulations using Ansys-CFX code have been done in steady state condition, characterizing the final phase of reflood. Peaks of temperature are observed in the middle of the fuel rod, followed by a temperature drop. This effect is justified by the increasing of heat transfer coefficient, originated from the high turbulence effects. Therefore, this paper considers a radial blockage of 90%, varying just the blockage length. This study observed that, for the same boundary conditions, the longer the blockage length originated after LOCA events, the higher are the central temperatures in the fuel rod. (author)

Safety analysis of RA reactor operation, I-III, Part II, Accident analysis

International Nuclear Information System (INIS)

Raisic, N.

1963-02-01

This volume covers the analyses of two types of accidents: accidents caused by uncontrolled reactivity increase, and accidents caused by decrease or loss of cooling. First type of accidents, uncontrolled reactivity insertion could occur due to removal of compensation, regulatory or safety rods, or by increase of heavy water level. Removal of irradiated samples from the core could also cause increase of reactivity. Second type of accidents could occur due to interruption of cooling, loss of water in the secondary cooling loop or loss of water in the primary coolant loop

Proton ejection project for Saturne; Projet d'ejection des protons de saturne

Energy Technology Data Exchange (ETDEWEB)

Bronca, G; Gendreau, G

1959-07-01

The reasons for choosing the ejection system are given. The characteristics required for the ejected beam are followed by a description of the ejection process, in chronological order from the viewpoint of the protons: movement of the particles, taking into account the various elements which make up the system (internal magnet, external magnet, quadrupoles, ejection correction coils, thin and thick cables,...) and specification of these elements. Then follows an estimation of the delay in manufacture and the cost of the project. Finally, the characteristics of the magnets and quadrupoles are listed in an appendix. (author) [French] On donne d'abord les raisons du choix du systeme d'ejection, puis le principe. Apres les caracteristiques requises pour le faisceau ejecte, on decrit le processus d'ejection selon l'ordre chronologique vu par les protons: mouvement des particules compte tenu des divers elements composant le systeme (aimant interne, aimant externe, quadrupoles, enroulements correcteurs ejection, cibles mince et epaisse,. ..) et cahier de charge de ces elements. On estime, ensuite les delais de realisation et le cout du projet. Enfin, un resume des caracteristiques des aimants et quadrupoles est donne en appendice. (auteur)

Commercial SNF Accident Release Fractions

Energy Technology Data Exchange (ETDEWEB)

J. Schulz

2004-11-05

The purpose of this analysis is to specify and document the total and respirable fractions for radioactive materials that could be potentially released from an accident at the repository involving commercial spent nuclear fuel (SNF) in a dry environment. The total and respirable release fractions are used to support the preclosure licensing basis for the repository. The total release fraction is defined as the fraction of total commercial SNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. Radionuclides are released from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses; this subset of the total release fraction is referred to as the respirable release fraction. Accidents may involve waste forms characterized as: (1) bare unconfined intact fuel assemblies, (2) confined intact fuel assemblies, or (3) canistered failed commercial SNF. Confined intact commercial SNF assemblies at the repository are contained in shipping casks, canisters, or waste packages. Four categories of failed commercial SNF are identified: (1) mechanically and cladding-penetration damaged commercial SNF, (2) consolidated/reconstituted assemblies, (3) fuel rods, pieces, and debris, and (4) nonfuel components. It is assumed that failed commercial SNF is placed into waste packages with a mesh screen at each end (CRWMS M&O 1999). In contrast to bare unconfined fuel assemblies, the

Commercial SNF Accident Release Fractions

International Nuclear Information System (INIS)

Schulz, J.

2004-01-01

The purpose of this analysis is to specify and document the total and respirable fractions for radioactive materials that could be potentially released from an accident at the repository involving commercial spent nuclear fuel (SNF) in a dry environment. The total and respirable release fractions are used to support the preclosure licensing basis for the repository. The total release fraction is defined as the fraction of total commercial SNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. Radionuclides are released from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses; this subset of the total release fraction is referred to as the respirable release fraction. Accidents may involve waste forms characterized as: (1) bare unconfined intact fuel assemblies, (2) confined intact fuel assemblies, or (3) canistered failed commercial SNF. Confined intact commercial SNF assemblies at the repository are contained in shipping casks, canisters, or waste packages. Four categories of failed commercial SNF are identified: (1) mechanically and cladding-penetration damaged commercial SNF, (2) consolidated/reconstituted assemblies, (3) fuel rods, pieces, and debris, and (4) nonfuel components. It is assumed that failed commercial SNF is placed into waste packages with a mesh screen at each end (CRWMS M andO 1999). In contrast to bare unconfined fuel assemblies, the

A thermal-hydraulic code for transient analysis in a channel with a rod bundle

International Nuclear Information System (INIS)

Khodjaev, I.D.

1995-01-01

The paper contains the model of transient vapor-liquid flow in a channel with a rod bundle of core of a nuclear power plant. The computer code has been developed to predict dryout and post-dryout heat transfer in rod bundles of nuclear reactor core under loss-of-coolant accidents. Economizer, bubble, dispersed-annular and dispersed regimes are taken into account. The computer code provides a three-field representation of two-phase flow in the dispersed-annular regime. Continuous vapor, continuous liquid film and entrained liquid drops are three fields. For the description of dispersed flow regime two-temperatures and single-velocity model is used. Relative droplet motion is taken into account for the droplet-to-vapor heat transfer. The conservation equations for each of regimes are solved using an effective numerical technique. This technique makes it possible to determine distribution of the parameters of flows along the perimeter of fuel elements. Comparison of the calculated results with the experimental data shows that the computer code adequately describes complex processes in a channel with a rod bundle during accident

A thermal-hydraulic code for transient analysis in a channel with a rod bundle

Energy Technology Data Exchange (ETDEWEB)

Khodjaev, I.D. [Research & Engineering Centre of Nuclear Plants Safety, Electrogorsk (Russian Federation)

1995-09-01

The paper contains the model of transient vapor-liquid flow in a channel with a rod bundle of core of a nuclear power plant. The computer code has been developed to predict dryout and post-dryout heat transfer in rod bundles of nuclear reactor core under loss-of-coolant accidents. Economizer, bubble, dispersed-annular and dispersed regimes are taken into account. The computer code provides a three-field representation of two-phase flow in the dispersed-annular regime. Continuous vapor, continuous liquid film and entrained liquid drops are three fields. For the description of dispersed flow regime two-temperatures and single-velocity model is used. Relative droplet motion is taken into account for the droplet-to-vapor heat transfer. The conservation equations for each of regimes are solved using an effective numerical technique. This technique makes it possible to determine distribution of the parameters of flows along the perimeter of fuel elements. Comparison of the calculated results with the experimental data shows that the computer code adequately describes complex processes in a channel with a rod bundle during accident.

Thermal hydraulic performance of naturally aspirated control rod housing assemblies

International Nuclear Information System (INIS)

Geiger, G.T.; Randolph, H.W.; Paik, I.K.; Foti, D.J.

1992-01-01

Savannah River Site reactors are comprised of heat generating fuel/target assemblies, control rods which regulate reactor power, and heavy water which acts as the coolant and as a moderator. The fuel/target assemblies are cooled by the downflow of heavy water while the control rods are cooled via upflow. Five control rods are grouped with two safety rods in seven-channel assemblies called septifoils. Under normal operating conditions, the reactor power level, radial shape flux and axial power flux are regulated by the positioning of the control rods. The control rods are solid rods of a lithium-aluminum alloy with an thin aluminum outer sheath. Lithium is a good absorber of neutrons and, thus control rod temperatures rise with reactor power. At conditions of sufficiently high reactor power and degraded coolant flow, the control rods could heat sufficiently to cause a metallurigical failure of the sheath leading to molten material coming in contact with water and the possibility of a steam explosion. An accident has been postulated as part of the analysis involving the safety upgrade of Savannah River Site reactors in which the housing is not seated on the pin. Coolant from the upflow pin would not be directed into the housing but, into the moderator space surrounding the housing. Only naturally aspirated cooling due to buoyancy effects would be available to cool the control rods and the coolant mass flow rate would drop significantly from its nominal value. In this study, the mechanisms and limits of cooling heated rods housed in an unseated septifoil are addressed. Experiments were conducted on a shortened, prototypic housing with electrically heated rods to gain an understanding of the phenomena governing the cooling in such a case and develop data which can be used to evaluate predictive models. These experiments are described, their results discussed, and the predictions of current models is presented

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Experimental study of the pressure discharge process for the hydraulic control rod drive system stepped cylinder

International Nuclear Information System (INIS)

Wang, Jinhua; Bo, Hanliang; Zheng, Wenxiang

2002-01-01

The pressure discharge process from the stepped cylinder of the Hydraulic Control Rod Drive System (HCRDS) was studied experimentally in the HCRDS experimental loop for the 200 MW Nuclear Heating Reactor (NHR-200). The results showed that the differential pressure between the outside and the inside of the stepped cylinder increased rapidly to the desired value so that the force induced by the differential pressure which pushes the out tube of stepped cylinder was large enough. Therefore, if the hydraulic control rod were jammed, the pressure could push the hydraulic control rod to overcome the frictional resistance to insert the control rod into the reactor core. The experimental results verified that this design would solve the problem of hydraulic control rod jamming during an accident. (author)

The droplet injection system used in the rod bundle heat transfer facility

International Nuclear Information System (INIS)

Frepoli, C.; Andrew, A.J.; Hochreiter, L.E.; Cheung, F.B.

2001-01-01

The full text follows. The US Nuclear Regulatory Commission (NRC) and the Pennsylvania State University are currently funding a research program entitled ''Rod Bundle Heat Transfer'' (RBHT). The main objective of the program is to investigate heat transfer during the core reflood period of a hypothetical Large Break Loss of Coolant Accident in a typical nuclear power plant. The RBHT test facility consists of a full-length 7 x 7 rod bundle. Information gathered by the RBHT test facility will be used for improvement of the reflood heat transfer models in the NRC's thermal hydraulic codes. In particular the RBHT data will be used to improve the understanding of individual heat transfer effects to the total rod heat transfer such that compensating errors present in current Best Estimate codes can be significantly reduced. The strategy in developing the test matrix is to use a ''building block'' approach in which simpler experiments are performed first to quantify a particular heat transfer mechanism alone and then the additional complications of the full two-phase flow, reflood film boiling behavior of the test facility are added in later experiments. One of these ''simpler'' experiments will be the injection of known size and velocity liquid droplets into the main stream of superheated steam. The droplet injection system consists of small diameter tubes inserted across the bundle at a given elevation. A number of equal size holes are drilled perpendicular to the surface in a triangular pitch. Water is forced into opposite ends of the tube and ejected from the holes. The injection system was tested using a digital imaging system known as VisiSizer. This system is capable of determining the diameter and velocity of small water droplets using a laser-illuminated digital camera system (LIDCS). Imaging software analyzes the digital images in real time to determine the distributions of droplet size and velocity. Pre-test analysis using COBRA-TF have been conducted to

SSYST, Modular System for Transient Fuel Rod Behaviour Under Accident Condition

International Nuclear Information System (INIS)

Gulden, W.; Meyder, R.; Borgwaldt, H.

1987-01-01

1 - Description of problem or function: SSYST is a code system for analyzing transient fuel rod behaviour under off-normal conditions, developed jointly by the Institut fuer Kernenergetik und Energie-systeme (IKE), Stuttgart, and Kernforschungszentrum Karlsruhe (KfK) under contract for the Projekt Nukleare Sicherheit (PNS) at KfK. Main differences versus codes with similar applications are: (1) an open-ended modular code organisation; (2) a preference for simple models, wherever possible. While feature (1) makes SSYST a very flexible tool, easily adapted to changing requirements, feature (2) leads to short execution times. The analysis of transient rod behaviour under LOCA boundary conditions takes 2 minutes CPU time on IBM 3033, so that extensive parametric studies are feasible. Main differences between SSYST-3 and previous versions are related to a general clean-up of the code system, which reduces the implementation effort: - advanced modules for cladding deformation and oxidation and reflooding conditions are included; - an input processor thoroughly checks all input data

A consolidation process for spent burnable poison rod assemblies

International Nuclear Information System (INIS)

Yamamoto, Y.; Harada, M.; Komatsu, Y.

1985-01-01

A new consolidation system for the spent burnable poison assembly utilizing a sequence control robot operated under water was proposed. A credible accident in the system was analyzed mainly from the viewpoint of tritium release, based on the diffusion analysis of tritium in borosilicate glass. It was found that the amount of tritium released would be small even after the rupture of burnable poison rods. An experiment on a new consolidation system was performed using spent burnable poison assemblies. The volume of burnable poison assemblies was reduced safely and securely by a factor of 7 to 14 for burnable poison rods and by 22 for hold-down portions. It was proved that the consolidation system is collectively feasible

Containment severe accident thermohydraulic phenomena

International Nuclear Information System (INIS)

Frid, W.

1991-08-01

This report describes and discusses the containment accident progression and the important severe accident containment thermohydraulic phenomena. The overall objective of the report is to provide a rather detailed presentation of the present status of phenomenological knowledge, including an account of relevant experimental investigations and to discuss, to some extent, the modelling approach used in the MAAP 3.0 computer code. The MAAP code has been used in Sweden as the main tool in the analysis of severe accidents. The dependence of the containment accident progression and containment phenomena on the initial conditions, which in turn are heavily dependent on the in-vessel accident progression and phenomena as well as associated uncertainties, is emphasized. The report is in three parts dealing with: * Swedish reactor containments, the severe accident mitigation programme in Sweden and containment accident progression in Swedish PWRs and BWRs as predicted by the MAAP 3.0 code. * Key non-energetic ex-vessel phenomena (melt fragmentation in water, melt quenching and coolability, core-concrete interaction and high temperature in containment). * Early containment threats due to energetic events (hydrogen combustion, high pressure melt ejection and direct containment heating, and ex-vessel steam explosions). The report concludes that our understanding of the containment severe accident progression and phenomena has improved very significantly over the parts ten years and, thereby, our ability to assess containment threats, to quantify uncertainties, and to interpret the results of experiments and computer code calculations have also increased. (au)

Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

International Nuclear Information System (INIS)

Arkoma, Asko; Hänninen, Markku; Rantamäki, Karin; Kurki, Joona; Hämäläinen, Anitta

2015-01-01

Highlights: • The number of failing fuel rods in a LB-LOCA in an EPR is evaluated. • 59 scenarios are simulated with the system code APROS. • 1000 rods per scenario are simulated with the fuel performance code FRAPTRAN-GENFLO. • All the rods in the reactor are simulated in the worst scenario. • Results suggest that the regulations set by the Finnish safety authority are met. - Abstract: In this paper, the number of failing fuel rods in a large break loss-of-coolant accident (LB-LOCA) in EPR-type nuclear power plant is evaluated using statistical methods. For this purpose, a statistical fuel failure analysis procedure has been developed. The developed method utilizes the results of nonparametric statistics, the Wilks’ formula in particular, and is based on the selection and variation of parameters that are important in accident conditions. The accident scenario is simulated with the coupled fuel performance – thermal hydraulics code FRAPTRAN-GENFLO using various parameter values and thermal hydraulic and power history boundary conditions between the simulations. The number of global scenarios is 59 (given by the Wilks’ formula), and 1000 rods are simulated in each scenario. The boundary conditions are obtained from a new statistical version of the system code APROS. As a result, in the worst global scenario, 1.2% of the simulated rods failed, and it can be concluded that the Finnish safety regulations are hereby met (max. 10% of the rods allowed to fail)

Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Arkoma, Asko, E-mail: asko.arkoma@vtt.fi; Hänninen, Markku; Rantamäki, Karin; Kurki, Joona; Hämäläinen, Anitta

2015-04-15

Highlights: • The number of failing fuel rods in a LB-LOCA in an EPR is evaluated. • 59 scenarios are simulated with the system code APROS. • 1000 rods per scenario are simulated with the fuel performance code FRAPTRAN-GENFLO. • All the rods in the reactor are simulated in the worst scenario. • Results suggest that the regulations set by the Finnish safety authority are met. - Abstract: In this paper, the number of failing fuel rods in a large break loss-of-coolant accident (LB-LOCA) in EPR-type nuclear power plant is evaluated using statistical methods. For this purpose, a statistical fuel failure analysis procedure has been developed. The developed method utilizes the results of nonparametric statistics, the Wilks’ formula in particular, and is based on the selection and variation of parameters that are important in accident conditions. The accident scenario is simulated with the coupled fuel performance – thermal hydraulics code FRAPTRAN-GENFLO using various parameter values and thermal hydraulic and power history boundary conditions between the simulations. The number of global scenarios is 59 (given by the Wilks’ formula), and 1000 rods are simulated in each scenario. The boundary conditions are obtained from a new statistical version of the system code APROS. As a result, in the worst global scenario, 1.2% of the simulated rods failed, and it can be concluded that the Finnish safety regulations are hereby met (max. 10% of the rods allowed to fail)

Energy analysis of control rod drive mechanism in HTR-10

International Nuclear Information System (INIS)

Bo Hanliang; Wu Yuanqiang

2000-01-01

This paper presents a theoretical model for the control rod drive mechanism for the 10 MW High Temperature Gas Cooled Reactor (HTR-10) and analyzes accidents which may occur in the drive mechanism, for example, chain break, coupling damage and other damage scenarios. The results show that the matching problem between buffer capability and coupling strength is the main reason for coupling damage; increased temperatures would reduce eddy damping and cause a mismatch between buffer capability and coupling strength; and the displacement of the buffer spring will affect the coupling force. The results provide a theoretical basis for the design of the control rod drive mechanism for HTR-10

Analysis of heat transfer from fuel rods with externally attached thermocouples

International Nuclear Information System (INIS)

Gill, C.R.; Coddington, P.

1988-05-01

This paper describes the development of 2 and 3 dimensional finite element heat conduction models to simulate the behaviour of the external thermocouples attached to the LOFT fuel rods during the blowdown phase of a large break loss-of-coolant accident. To establish the model and determine the thermal coupling between the thermocouple and the fuel rod extensive use was made of two series of experiments performed at INEL in the LOFT Test Support Facility (LTSF). These experiments were high pressure reflood experiments with fluid conditions 'typical' of those seen during the bottom-up flow period of the LOFT experiments. (author)

Core loss during a severe accident (COLOSS)

International Nuclear Information System (INIS)

Adroguer, B.; Bertrand, F.; Chatelard, P.; Cocuaud, N.; Van Dorsselaere, J.P.; Bellenfant, L.; Knocke, D.; Bottomley, D.; Vrtilkova, V.; Belovsky, L.; Mueller, K.; Hering, W.; Homann, C.; Krauss, W.; Miassoedov, A.; Schanz, G.; Steinbrueck, M.; Stuckert, J.; Hozer, Z.; Bandini, G.; Birchley, J.; Berlepsch, T. von; Kleinhietpass, I.; Buck, M.; Benitez, J.A.F.; Virtanen, E.; Marguet, S.; Azarian, G.; Caillaux, A.; Plank, H.; Boldyrev, A.; Veshchunov, M.; Kobzar, V.; Zvonarev, Y.; Goryachev, A.

2005-01-01

The COLOSS project was a 3-year shared-cost action, which started in February 2000. The work-programme performed by 19 partners was shaped around complementary activities aimed at improving severe accident codes. Unresolved risk-relevant issues regarding H 2 production, melt generation and the source term were studied through a large number of experiments such as (a) dissolution of fresh and high burn-up UO 2 and MOX by molten Zircaloy (b) simultaneous dissolution of UO 2 and ZrO 2 (c) oxidation of U-O-Zr mixtures (d) degradation-oxidation of B 4 C control rods. Corresponding models were developed and implemented in severe accident computer codes. Upgraded codes were then used to apply results in plant calculations and evaluate their consequences on key severe accident sequences in different plants involving B 4 C control rods and in the TMI-2 accident. Significant results have been produced from separate-effects, semi-global and large-scale tests on COLOSS topics enabling the development and validation of models and the improvement of some severe accident codes. Breakthroughs were achieved on some issues for which more data are needed for consolidation of the modelling in particular on burn-up effects on UO 2 and MOX dissolution and oxidation of U-O-Zr and B 4 C-metal mixtures. There was experimental evidence that the oxidation of these mixtures can contribute significantly to the large H 2 production observed during the reflooding of degraded cores under severe accident conditions. The plant calculation activity enabled (a) the assessment of codes to calculate core degradation with the identification of main uncertainties and needs for short-term developments and (b) the identification of safety implications of new results. Main results and recommendations for future R and D activities are summarized in this paper

Comprehensive and consistent interpretation of local fault experiments and application to hypothetical local overpower accident in Monju

International Nuclear Information System (INIS)

Fukano, Yoshitaka

2013-01-01

Experimental studies on local fault (LF) accidents in fast breeder reactors have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Comprehensive and consistent interpretations of in-pile and out-of-pile experiments related to LF were arrived at in this study based on state-of-the-art review and data analysis techniques. Safety margins for a hypothetical local overpower accident, which was evaluated as a LF accident in the licensing document of the construction permit for a prototype fast breeder reactor called Monju, were also studied. Based on comprehensive interpretations of the latest experimental database, including those performed after the permission of Monju construction, it was clarified that the evaluation of the hypothetical local overpower accident in the Monju licensing was sufficiently conservative. Furthermore, it incorporated adequate safety margins in terms of failure thresholds of the fuel pin, molten fuel ejection, fuel sweep-out behavior after molten fuel ejection, and pin-to-pin failure propagation. Moreover, these comprehensive interpretations are valid and applicable to the safety evaluation of LF accidents of other fast breeder reactors with various fuel and core designs. (author)

Progress on B4C control rod modeling in RELAP/SCDAPSIM with application to quench and Phebus

International Nuclear Information System (INIS)

Kawahara, Keisuke; Hohorst, Judith K.; Allison, Chris M.

2014-01-01

The RELAP/SCDAPSIM code is designed to predict the behavior of reactor systems during normal and accident conditions. RELAP/SCDAPSIM/MOD3.5 is an experimental version of the code with the most advanced fuel and severe accident behavior models and correlations. It includes modeling improvements that were specifically added to support (a) the ongoing experimental severe accident programs in Europe and Japan and (b) the analysis and assessment activities related to the accident at the Fukushima Daiichi NPS. One of the improved models describes the behavior of cylindrical B 4 C control rods used in selected PWR designs and in integral experiments used to assess the heating and melting of PWR, BWR, and VVER assemblies. It replaces an older model that was originally developed by the US Nuclear Regulatory Commission in the mid- 1980's. It includes a combination of new and improved models and correlations to more accurately describe (a) eutectic reactions between Zircaloy, B 4 C, and stainless steel, (b) oxidation for B 4 C, Zircaloy, and stainless steel, and (c) the effects of the gap between the Zircaloy guide tube and the stainless steel sheath surrounding B 4 C pellets used in many control rod designs. This paper will discuss the development of the new model and validation of the model using the PHEBUS B 4 C test, FPT-3, and the KIT quench experiments with a central B 4 C control rod. (authors)

B4C control rod behavior during severe accident sequences

International Nuclear Information System (INIS)

Steinbrueck, M.

2003-01-01

The oxidation kinetics of various types of boron carbides (pellets, powder) as well as the degradation of B 4 C control rod segments were investigated in the temperature range between 800 and 1600 deg C. Mass spectrometric gas analysis was used to determine oxidation rates in transient and isothermal tests. The oxidation kinetics of boron carbide are determined by the formation of a liquid boron oxide layer and its loss due to the reaction with surplus steam to form volatile boric acids and at temperatures above 1500 deg C by direct evaporation. Under these test conditions linear oxidation kinetics are established soon after oxidation has initiated. The oxidation kinetics are strongly influenced by the thermal-hydraulic boundary conditions, in particular by the steam flow rate. Only very low amounts of methane were ever produced in these tests. Enhanced degradation of B 4 C control rods starts with the rapid formation of eutectic melts in the systems B 4 C-stainless steel (SS) and SS-Zircaloy at temperatures above 1250 deg C. Initially, this melt is kept within a ZrO 2 scale externally formed at the Zircaloy guide tube. The absorber melt is rapidly oxidized after failure of the oxide shell and aggressively attacks adjacent fuel claddings. (author)

Investigation of Swirling Flow in Rod Bundle Subchannels Using Computational Fluid Dynamics

International Nuclear Information System (INIS)

Holloway, Mary V.; Beasley, Donald E.; Conner, Michael E.

2006-01-01

The fluid dynamics for turbulent flow through rod bundles representative of those used in pressurized water reactors is examined using computational fluid dynamics (CFD). The rod bundles of the pressurized water reactor examined in this study consist of a square array of parallel rods that are held on a constant pitch by support grids spaced axially along the rod bundle. Split-vane pair support grids are often used to create swirling flow in the rod bundle in an effort to improve the heat transfer characteristics for the rod bundle during both normal operating conditions and in accident condition scenarios. Computational fluid dynamics simulations for a two subchannel portion of the rod bundle were used to model the flow downstream of a split-vane pair support grid. A high quality computational mesh was used to investigate the choice of turbulence model appropriate for the complex swirling flow in the rod bundle subchannels. Results document a central swirling flow structure in each of the subchannels downstream of the split-vane pairs. Strong lateral flows along the surface of the rods, as well as impingement regions of lateral flow on the rods are documented. In addition, regions of lateral flow separation and low axial velocity are documented next to the rods. Results of the CFD are compared to experimental particle image velocimetry (PIV) measurements documenting the lateral flow structures downstream of the split-vane pairs. Good agreement is found between the computational simulation and experimental measurements for locations close to the support grid. (authors)

Parametric Evaluation of SiC/SiC Composite Cladding with UO2 Fuel for LWR Applications: Fuel Rod Interactions and Impact of Nonuniform Power Profile in Fuel Rod

Science.gov (United States)

Singh, G.; Sweet, R.; Brown, N. R.; Wirth, B. D.; Katoh, Y.; Terrani, K.

2018-02-01

SiC/SiC composites are candidates for accident tolerant fuel cladding in light water reactors. In the extreme nuclear reactor environment, SiC-based fuel cladding will be exposed to neutron damage, significant heat flux, and a corrosive environment. To ensure reliable and safe operation of accident tolerant fuel cladding concepts such as SiC-based materials, it is important to assess thermo-mechanical performance under in-reactor conditions including irradiation and realistic temperature distributions. The effect of non-uniform dimensional changes caused by neutron irradiation with spatially varying temperatures, along with the closing of the fuel-cladding gap, on the stress development in the cladding over the course of irradiation were evaluated. The effect of non-uniform circumferential power profile in the fuel rod on the mechanical performance of the cladding is also evaluated. These analyses have been performed using the BISON fuel performance modeling code and the commercial finite element analysis code Abaqus. A constitutive model is constructed and solved numerically to predict the stress distribution in the cladding under normal operating conditions. The dependence of dimensions and thermophysical properties on irradiation dose and temperature has been incorporated into the models. Initial scoping results from parametric analyses provide time varying stress distributions in the cladding as well as the interaction of fuel rod with the cladding under different conditions of initial fuel rod-cladding gap and linear heat rate. It is found that a non-uniform circumferential power profile in the fuel rod may cause significant lateral bowing in the cladding, and motivates further analysis and evaluation.

Effects of fuel relocation on reflood in a partially-blocked rod bundle

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Jae [School of Mechanical Engineering, Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon 34134 (Korea, Republic of); Kim, Jongrok; Kim, Kihwan; Bae, Sung Won [Thermal-Hydraulic Safety Research Division, Korea Atomic Energy Research Division, 111 Daedeok-daero, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Moon, Sang-Ki, E-mail: skmoon@kaeri.re.kr [Thermal-Hydraulic Safety Research Division, Korea Atomic Energy Research Division, 111 Daedeok-daero, Yuseong-gu, Daejeon 34057 (Korea, Republic of)

2017-02-15

Ballooning of the fuel rods has been an important issue, since it can influence the coolability of the rod bundle in a large-break loss-of-coolant accident (LBLOCA). Numerous past studies have investigated the effect of blockage geometry on the heat transfer in a partially blocked rod bundle. However, they did not consider the occurrence of fuel relocation and the corresponding effect on two-phase heat transfer. Some fragmented fuel particles located above the ballooned region may drop into the enlarged volume of the balloon. Accordingly, the fuel relocation brings in a local power increase in the ballooned region. The present study’s objective is to investigate the effect of the fuel relocation on the reflood under a LBLOCA condition. Toward this end, experiments were performed in a 5 × 5 partially-blocked rod bundle. Two power profiles were tested: one is a typical cosine shape and the other is the modified shape considering the effect of the fuel relocation. For a typical power shape, the peak temperature in the ballooned rods was lower than that in the intact rods. On the other hand, for the modified power shape, the peak temperature in the ballooned rods was higher than that in the intact rods. Numerical simulations were also performed using the MARS code. The tendencies of the peak clad temperatures were well predicted.

Ejection Tower Lab

Data.gov (United States)

Federal Laboratory Consortium — The Ejection Tower Facility's mission is to test and evaluate new ejection seat technology being researched and developed for future defense forces. The captive and...

LOFT fuel rod surface temperature measurement testing

International Nuclear Information System (INIS)

Eaton, A.M.; Tolman, E.L.; Solbrig, C.W.

1978-01-01

Testing of the LOFT fuel rod cladding surface thermocouples has been performed to evaluate how accurately the LOFT thermocouples measure the cladding surface temperature during a loss-of-coolant accident (LOCA) sequence and what effect, if any, the thermocouple would have on core performance. Extensive testing has been done to characterize the thermocouple design. Thermal cycling and corrosion testing of the thermocouple weld design have provided an expected lifetime of 6000 hours when exposed to reactor coolant conditions of 620 K and 15.9 MPa and to sixteen thermal cycles with an initial temperature of 480 K and peak temperatures ranging from 870 to 1200K. Departure from nucleate boiling (DNB) tests have indicated a DNB penalty (5 to 28% lower) during steady state operation and negligible effects during LOCA blowdown caused by the LOFT fuel rod surface thermocouple arrangement. Experience with the thermocouple design in Power Burst Facility (PBF) and LOFT nonnuclear blowdown testing has been quite satisfactory. Tests discussed here were conducted using both stainless steel and zircaloy-clad electrically heated rod in the LOFT Test Support Facility (LTSF) blowdown simulation loop

Utilization of control rod drive (CRD) system for long term core cooling

International Nuclear Information System (INIS)

Huerta B, A.

1991-01-01

In this paper we consider an application of Probabilistic Risk Assessment (PRA) to risk management. Foreseeable risk management strategies to prevent core damage are constrained by the availability of first line systems as well as support systems. The actual trend in the evaluation of risk management options can be performed in a number of ways. An example is the identification of back-up systems which could be used to perform the same safety functions. In this work we deal with the evaluation of the feasibility, for BWR's, to use the Control Rod Drive system to maintain an adequate reactor core long term cooling in some accident sequences. This preliminary evaluation is carried out as a part of the Internal Events Analysis for Laguna Verde Nuclear Power Plant (LVNPP) that is currently under way by the Mexican Nuclear Regulatory Body. This analysis addresses the evaluation and incorporation of all the systems, including the safety related and the back-up non safety related systems, that are available for the operator in order to prevent core damage. As a part of this analysis the containment venting capability is also evaluated as a back-up of the containment heat removal function. This will prevent the primary containment overpressurization and loss of certain core cooling systems. A selection of accident sequences in which the Control Rod Drive system could be used to mitigate the accident and prevent core damage are discussed. A personal computer transient analysis code is used to carry out thermohydraulic simulations in order to evaluate the Control Rod Drive system performance, the corresponding results are presented. Finally, some preliminary conclusions are drawn. (author). 9 refs, 5 figs

COMMERCIAL SNF ACCIDENT RELEASE FRACTIONS

Energy Technology Data Exchange (ETDEWEB)

S.O. Bader

1999-10-18

The purpose of this design analysis is to specify and document the total and respirable fractions for radioactive materials that are released from an accident event at the Monitored Geologic Repository (MGR) involving commercial spent nuclear fuel (CSNF) in a dry environment. The total and respirable release fractions will be used to support the preclosure licensing basis for the MGR. The total release fraction is defined as the fraction of total CSNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. The radionuclides are released from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses. This subset of the total release fraction is referred to as the respirable release fraction. Potential accidents may involve waste forms that are characterized as either bare (unconfined) fuel assemblies or confined fuel assemblies. The confined CSNF assemblies at the MGR are contained in shipping casks, canisters, or disposal containers (waste packages). In contrast to the bare fuel assemblies, the container that confines the fuel assemblies has the potential of providing an additional barrier for diminishing the total release fraction should the fuel rod cladding breach during an accident. However, this analysis will not take credit for this additional bamer and will establish only the total release fractions for bare unconfined CSNF assemblies, which may however be

COMMERCIAL SNF ACCIDENT RELEASE FRACTIONS

International Nuclear Information System (INIS)

S.O. Bader

1999-01-01

The purpose of this design analysis is to specify and document the total and respirable fractions for radioactive materials that are released from an accident event at the Monitored Geologic Repository (MGR) involving commercial spent nuclear fuel (CSNF) in a dry environment. The total and respirable release fractions will be used to support the preclosure licensing basis for the MGR. The total release fraction is defined as the fraction of total CSNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. The radionuclides are released from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses. This subset of the total release fraction is referred to as the respirable release fraction. Potential accidents may involve waste forms that are characterized as either bare (unconfined) fuel assemblies or confined fuel assemblies. The confined CSNF assemblies at the MGR are contained in shipping casks, canisters, or disposal containers (waste packages). In contrast to the bare fuel assemblies, the container that confines the fuel assemblies has the potential of providing an additional barrier for diminishing the total release fraction should the fuel rod cladding breach during an accident. However, this analysis will not take credit for this additional bamer and will establish only the total release fractions for bare unconfined CSNF assemblies, which may however be

Transient debris freezing and potential wall melting during a severe reactivity initiated accident experiment

International Nuclear Information System (INIS)

El-Genk, M.S.; Moore, R.L.

1981-01-01

It is important to light water reactor (LWR) safety analysis to understand the transient freezing of molten core debris on cold structures following a hypothetical core meltdown accident. The purpose of this paper is to (a) present the results of a severe reactivity initiated accident (RIA) in-pile experiment with regard to molten debris distribution and freezing following test fuel rod failure, (b) analyze the transient freezing of molten debris (primarily a mixture of UO/sub 2/ fuel and Zircaloy cladding) deposited on the inner surface of the test shroud wall upon rod failure, and (c) assess the potential for wall melting upon being contacted by the molten debris. 26 refs

Control rod

International Nuclear Information System (INIS)

Kawakami, Kazuo; Shimoshige, Takanori; Nishimura, Akira

1979-01-01

Purpose: A control rod has been developed, which provided a plurality of through-holes in the vicinity of the sheath fitting position, in order to flatten burn-up, of fuel rods in positions confronting a control rod. Thereby to facilitate the manufacture of the control rods and prevent fuel rod failures. Constitution: A plurality of through-holes are formed in the vicinity of the sheath fitting position of a central support rod to which a sheath for the control rod is fitted. These through-holes are arranged in the axial direction of the central support rod. Accordingly, burn-up of fuel rods confronting the control rods can be reduced by through-holes and fuel rod failures can be prevented. (Yoshino, Y.)

Release of fission products during controlled loss-of-coolant accidents and hypothetical core meltdown accidents

International Nuclear Information System (INIS)

Albrecht, H.; Malinauskas, A.P.

1978-01-01

A few years ago the Projekt Nukleare Sicherheit joined the United States Nuclear Regulatory Commission in the development of a research program which was designed to investigate fission product release from light water reactor fuel under conditions ranging from spent fuel shipping cask accidents to core meltdown accidents. Three laboratories have been involved in this cooperative effort. At Argonne National Laboratory (ANL), the research effort has focused on noble gas fission product release, whereas at Oak Ridge National Laboratory (ORNL) and at Kernforschungszentrum Karlsruhe (KfK), the studies have emphasized the release of species other than the noble gases. In addition, the ORNL program has been directed toward the development of fission product source terms applicable to analyses of spent fuel shipping cask accidents and controlled loss-of-coolant accidents, and the KfK program has been aimed at providing similar source terms which are characteristic of core meltdown accidents. The ORNL results are presented for fission product release from defected fuel rods into a steam atmosphere over the temperature range 500 to 1200 0 C, and the KfK results for release during core meltdown sequences

Transients analysis able to lead Pressurised Water Reactors cores to degraded situations, analysis of resulting configurations

International Nuclear Information System (INIS)

Shin, Hyeong-Ki

1999-01-01

The severe accidents that occurred recently on nuclear reactors such as Chernobyl and T.M.1.2 have led many countries utilizing nuclear energy to examine their severe accident management. This thesis focuses on this problem and aims at analyzing, in terms of reactivity, degraded core behavior resulting from different accidental configurations. Two types of core degradation can be encountered: local degradation (the destruction of isolated assemblies in the core) or spreading degradation (the destruction of neighboring assemblies). The TMI accident is an example of spreading degradation in the core. The simplicity of implementing the control rod ejection accident calculation as compared to other accidental transients have motivated the choice of this accident as a determinant for local degraded core configurations. The control rod ejection accident presents important three dimensional effects and introduces neutronic/thermohydraulic coupling. The implementation and validation of already existing three dimensional coupled calculation scheme, allowed one to analyze the consequences of such an accident and to the conclusion that only unrealistic hypotheses of assembly permutation could lead to a partial core degradation. A reasonable estimate of stored energy in the assemblies with high bum up, in relation to the stored energy in the hot spot, was also obtained for the first time. The recently performed experiments (CABRI experiments) showed that in highly burned up assemblies, the capacity to store energy decreases strongly in relation to new assemblies. This first estimate of the distribution of produced energy between different assemblies, during the rod ejection accident, offers an important piece of knowledge in the study of the consequences of an eventual fuel cycle extension (presently under consideration by development companies). Finally, the analysis of degraded core reactivity itself has been performed for a vast range of the degraded core configurations

The experimental development and performance test of the pneumatic control-rod drive for the THTR

International Nuclear Information System (INIS)

Lange, G.; Boehlo, D.; Heim, H.; Kleine-Tebbe, A.

1976-01-01

Reactor control and shutdown of the THTR is accomplished by two independent systems, the first consisting of 36 absorber rods penetrating the graphite reflector region surrounding the core, the second consisting of 42 absorber rods that insert directly into the pebble bed core. This paper describes the design development and testing of the pneumatic rod drives used for movement of the 42 core control rods. The core control rods have two functions: the first, for reactor safety purposes, provides for adequate safe shutdown of the reactor under cold conditions; the second, for operational purposes, provides for compensation of slow changes in reactivity. The safety and operational functions for each absorber rod are respectively carried out by a long-stroke-piston pneumatic drive and by a stepping-piston pneumatic drive, both of these independent, helium-driven drives being incorporated in the rod drive unit for each control rod. To study the performance of the rod drive, a complete prototype control rod and rod drive unit was built and tested under simulated reactor operational conditions. Operational experience under helium temperatures and pressures was gained and the drives were tested under stress and simulated accident conditions. The reliability of this system has been demonstrated to licensing authorities and to the customer. The programme will be completed with the commissioning tests of drives for the THTR-300 reactor. (author)

Simulation of fuel rod behaviour during various break LOCAs in PWRs

International Nuclear Information System (INIS)

Gadalla, A.A.; El-Fawal, M.M.

1996-01-01

During loss of coolant accident (LOCAs) course of events, attention focuses on fuel rod cladding temperature behaviour. In this study, the DRUFAN analytical model and LOBI-MOD2 experimental modeling scheme for fuel rod temperature behaviour during C L-Break LOCA in PWRs, are described and discussed. These models are applied for the investigation of fuel rod cladding temperature behaviour during LOCA blowdown phase. A spectrum of selected values representing small, intermediate and large CL- Break sizes are considered in the predictions. The results of the predictions demonstrated that calculated heater rod temperature at steady state as well as the transient period up to 1000 sec are going in good agreement with the measured values. However above 1000 sec the calculated temperatures are higher than the measured values. This indicates that code predictions in this period are conservative. The results indicated also that, in case of small CL-break LOCA (0.01 A and 0.01 and 0.03 A), the heater rod cladding temperature don't rise above saturation temperature. However, on the top of the heater rod, DNB is occurred in case of 0.03 A CL break, while for 0.01 A break, DNB didn't occur. In case of intermediate and large CL-break; (0.05 A, 0.10 A and 1 A), the results showed that, the heater rod cladding temperature exceeded the saturation temperature and DNB prevailed in upper and intermediate sections of the core. 15 figs., 2 tabs

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

International Nuclear Information System (INIS)

Parsons, P.D.; Mowat, J.A.S.; Dewhurst, D.W.F.; Hughes, T.E.

1983-01-01

An experimental study of the interaction between Zircaloy-4 cladding and UO 2 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 UO 2 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)

3-D core modelling of RIA transient: the TMI-1 benchmark

International Nuclear Information System (INIS)

Ferraresi, P.; Studer, E.; Avvakumov, A.; Malofeev, V.; Diamond, D.; Bromley, B.

2001-01-01

The increase of fuel burn up in core management poses actually the problem of the evaluation of the deposited energy during Reactivity Insertion Accidents (RIA). In order to precisely evaluate this energy, 3-D approaches are used more and more frequently in core calculations. This 'best-estimate' approach requires the evaluation of code uncertainties. To contribute to this evaluation, a code benchmark has been launched. A 3-D modelling for the TMI-1 central Ejected Rod Accident with zero and intermediate initial powers was carried out with three different methods of calculation for an inserted reactivity respectively fixed at 1.2 $ and 1.26 $. The studies implemented by the neutronics codes PARCS (BNL) and CRONOS (IPSN/CEA) describe an homogeneous assembly, whereas the BARS (KI) code allows a pin-by-pin representation (CRONOS has both possibilities). All the calculations are consistent, the variation in figures resulting mainly from the method used to build cross sections and reflectors constants. The maximum rise in enthalpy for the intermediate initial power (33 % P N ) calculation is, for this academic calculation, about 30 cal/g. This work will be completed in a next step by an evaluation of the uncertainty induced by the uncertainty on model parameters, and a sensitivity study of the key parameters for a peripheral Rod Ejection Accident. (authors)

3-D core modelling of RIA transient: the TMI-1 benchmark

Energy Technology Data Exchange (ETDEWEB)

Ferraresi, P. [CEA Cadarache, Institut de Protection et de Surete Nucleaire, Dept. de Recherches en Securite, 13 - Saint Paul Lez Durance (France); Studer, E. [CEA Saclay, Dept. Modelisation de Systemes et Structures, 91 - Gif sur Yvette (France); Avvakumov, A.; Malofeev, V. [Nuclear Safety Institute of Russian Research Center, Kurchatov Institute, Moscow (Russian Federation); Diamond, D.; Bromley, B. [Nuclear Energy and Infrastructure Systems Div., Brookhaven National Lab., BNL, Upton, NY (United States)

2001-07-01

The increase of fuel burn up in core management poses actually the problem of the evaluation of the deposited energy during Reactivity Insertion Accidents (RIA). In order to precisely evaluate this energy, 3-D approaches are used more and more frequently in core calculations. This 'best-estimate' approach requires the evaluation of code uncertainties. To contribute to this evaluation, a code benchmark has been launched. A 3-D modelling for the TMI-1 central Ejected Rod Accident with zero and intermediate initial powers was carried out with three different methods of calculation for an inserted reactivity respectively fixed at 1.2 $ and 1.26 $. The studies implemented by the neutronics codes PARCS (BNL) and CRONOS (IPSN/CEA) describe an homogeneous assembly, whereas the BARS (KI) code allows a pin-by-pin representation (CRONOS has both possibilities). All the calculations are consistent, the variation in figures resulting mainly from the method used to build cross sections and reflectors constants. The maximum rise in enthalpy for the intermediate initial power (33 % P{sub N}) calculation is, for this academic calculation, about 30 cal/g. This work will be completed in a next step by an evaluation of the uncertainty induced by the uncertainty on model parameters, and a sensitivity study of the key parameters for a peripheral Rod Ejection Accident. (authors)

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

Energy Technology Data Exchange (ETDEWEB)

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)

Porosity effects during a severe accident

International Nuclear Information System (INIS)

Cazares R, R. I.; Espinosa P, G.; Vazquez R, A.

2015-09-01

The aim of this work is to study the behaviour of porosity effects on the temporal evolution of the distributions of hydrogen concentration and temperature profiles in a fuel assembly where a stream of steam is flowing. The analysis considers the fuel element without mitigation effects. The mass transfer phenomenon considers that the hydrogen generated diffuses in the steam by convection and diffusion. Oxidation of the cladding, rods and other components in the core constructed in zirconium base alloy by steam is a critical issue in LWR accident producing severe core damage. The oxygen consumed by the zirconium is supplied by the up flow of steam from the water pool below the uncovered core, supplemented in the case of PWR by gas recirculation from the cooler outer regions of the core to hotter zones. Fuel rod cladding oxidation is then one of the key phenomena influencing the core behavior under high-temperature accident conditions. The chemical reaction of oxidation is highly exothermic, which determines the hydrogen rate generation and the cladding brittleness and degradation. The heat transfer process in the fuel assembly is considered with a reduced order model. The Boussinesq approximation was applied in the momentum equations for multicomponent flow analysis that considers natural convection due to buoyancy forces, which is related with thermal and hydrogen concentration effects. The numerical simulation was carried out in an averaging channel that represents a core reactor with the fuel rod with its gap and cladding and cooling steam of a BWR. (Author)

Porosity effects during a severe accident

Energy Technology Data Exchange (ETDEWEB)

Cazares R, R. I. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Posgrado en Energia y Medio Ambiente, San Rafael Atlixco 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Espinosa P, G.; Vazquez R, A., E-mail: ricardo-cazares@hotmail.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, San Rafael Atlixco 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico)

2015-09-15

The aim of this work is to study the behaviour of porosity effects on the temporal evolution of the distributions of hydrogen concentration and temperature profiles in a fuel assembly where a stream of steam is flowing. The analysis considers the fuel element without mitigation effects. The mass transfer phenomenon considers that the hydrogen generated diffuses in the steam by convection and diffusion. Oxidation of the cladding, rods and other components in the core constructed in zirconium base alloy by steam is a critical issue in LWR accident producing severe core damage. The oxygen consumed by the zirconium is supplied by the up flow of steam from the water pool below the uncovered core, supplemented in the case of PWR by gas recirculation from the cooler outer regions of the core to hotter zones. Fuel rod cladding oxidation is then one of the key phenomena influencing the core behavior under high-temperature accident conditions. The chemical reaction of oxidation is highly exothermic, which determines the hydrogen rate generation and the cladding brittleness and degradation. The heat transfer process in the fuel assembly is considered with a reduced order model. The Boussinesq approximation was applied in the momentum equations for multicomponent flow analysis that considers natural convection due to buoyancy forces, which is related with thermal and hydrogen concentration effects. The numerical simulation was carried out in an averaging channel that represents a core reactor with the fuel rod with its gap and cladding and cooling steam of a BWR. (Author)

Coupled neutronic and thermal-hydraulic code benchmark activities at the International Nuclear Safety Center

International Nuclear Information System (INIS)

Podlazov, L. N.

1998-01-01

Two realistic benchmark problems are defined and used to assess the performance of coupled thermal-hydraulic and neutronic codes used in simulating dynamic processes in VVER-1000 and RBMK reactor systems. One of the problems simulates a design basis accident involving the ejection of three control and protection system rods from a VVER-1000 reactor. The other is based on a postulated rod withdrawal from an operating RBMK reactor. Preliminary results calculated by various codes are compared. While these results show significant differences, the intercomparisons performed so far provide a basis for further evaluation of code limitations and modeling assumptions

Control rod

International Nuclear Information System (INIS)

Igarashi, Takao; Sugawara, Satoshi; Yoshimoto, Yuichiro; Saito, Shozo; Fukumoto, Takashi.

1987-01-01

Purpose: To reduce the weight and thereby obtain satisfactory operationability of control rods by combining absorbing nuclear chain type neutron absorbers and conventional type neutron absorbers in the axial direction of blades. Constitution: Neutron absorber rods and long life type neutron absorber rods are disposed in a tie rod and a sheath. The neutron absorber rod comprises a poison tube made of stainless steels and packed with B 4 C powder. The long life type neutron absorber rod is prepared by packing B-10 enriched boron carbide powder into a hafnium metal rod, hafnium pipe, europium and stainless made poison tube. Since the long life type absorber rod uses HF as the absorbing nuclear chain type neutron absorber, it absorbs neutrons to form new neutron absorbers to increase the nuclear life. (Yoshino, Y.)

Control rods

International Nuclear Information System (INIS)

Maruyama, Hiromi.

1984-01-01

Purpose: To realize effective utilization, cost reduction and weight reduction in neutron absorbing materials. Constitution: Residual amount of neutron absorbing material is averaged between the top end region and other regions of a control rod upon reaching to the control rod working life, by using a single kind of neutron absorbing material and increasing the amount of the neutron absorber material at the top end region of the control rod as compared with that in the other regions. Further, in a case of a control rod having control rod blades such as in a cross-like control rod, the amount of the neutron absorbing material is decreased in the middle portion than in the both end portions of the control rod blade along the transversal direction of the rod, so that the residual amount of the neutron absorbing material is balanced between the central region and both end regions upon reaching the working life of the control rod. (Yoshihara, H.)

Deformation of PWR cladding following a loss-of-coolant accident

International Nuclear Information System (INIS)

Hindle, E.D.; Mann, C.A.

1979-07-01

A review is presented of recent experiments to investigate the deformation behaviour of Zircaloy cladding in simulated loss-of-coolant accidents. The behaviour of Zircaloy cladding is shown to be controlled by a complex interaction of metallurgical and heat transfer variables, with the latter having a major influence. There is a significant increase in both diametral strain and the axial extent of deformation in multi-rod compared with single-rod tests. The extent to which this will occur in nuclear-heated tests is not yet known; however, it is expected that the 'smearing' of the gamma-radiation portion of decay heat in such tests will tend to reduce circumferential temperature variations. Opposing this is the influence of the colder control rods in an assembly. The resolution of this dichotomy will require a series of in-reactor multi-rod tests and attendant code development. (author)

Control rod calibration including the rod coupling effect

International Nuclear Information System (INIS)

Szilard, R.; Nelson, G.W.

1984-01-01

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

Determination of the control rod worth for research reactors

International Nuclear Information System (INIS)

Aldama, D.L.; Gual, M.R.

2000-01-01

Nowadays there is a big interest in developing neutronic analysis methods for research reactor and particularly for the determination of the control rods worth under different operation conditions and core configurations. The reactivity associated with the control rods is of interest in the shutdown margin and in calculations of possible abnormal conditions related to reactivity accidents. For theses studies several computer codes have been developed. The present work is aimed at the validation of the calculation methods of the Nuclear Technology Center of Cuba. For this purpose, in order to evaluate the safety of this type of installations, the reactivity worth of the control rods of the cylindrical configuration of the Brazilian critical assembly IPEN/MB-01 is determined. These calculations, however, are a relatively complex task that requires the use of three-dimensional models. Because of this, the validation of the calculation methods used for this purpose is of great importance. In fact, it is one of the requirements called upon by the quality assurance programs for the development, maintenance and utilization of the calculation codes used in safety analysis. For the calculation of control rod worth the lattice code WIMS-D/4 [8] and the diffusion code SNAP-3D [9] were used. This work presents the obtained results and gives a comparison with the experimental values

Control Rod Driveline Reactivity Feedback Model for Liquid Metal Reactors

International Nuclear Information System (INIS)

Kwon, Young-Min; Jeong, Hae-Yong; Chang, Won-Pyo; Cho, Chung-Ho; Lee, Yong-Bum

2008-01-01

The thermal expansion of the control rod drivelines (CRDL) is one important passive mitigator under all unprotected accident conditions in the metal and oxide cores. When the CRDL are washed by hot sodium in the coolant outlet plenum, the CRDL thermally expands and causes the control rods to be inserted further down into the active core region, providing a negative reactivity feedback. Since the control rods are attached to the top of the vessel head and the core attaches to the bottom of the reactor vessel (RV), the expansion of the vessel wall as it heats will either lower the core or raise the control rods supports. This contrary thermal expansion of the reactor vessel wall pulls the control rods out of the core somewhat, providing a positive reactivity feedback. However this is not a safety factor early in a transient because its time constant is relatively large. The total elongated length is calculated by subtracting the vessel expansion from the CRDL expansion to determine the net control rod expansion into the core. The system-wide safety analysis code SSC-K includes the CRDL/RV reactivity feedback model in which control rod and vessel expansions are calculated using single-nod temperatures for the vessel and CRDL masses. The KALIMER design has the upper internal structures (UIS) in which the CRDLs are positioned outside the structure where they are exposed to the mixed sodium temperature exiting the core. A new method to determine the CRDL expansion is suggested. Two dimensional hot pool thermal hydraulic model (HP2D) originally developed for the analysis of the stratification phenomena in the hot pool is utilized for a detailed heat transfer between the CRDL mass and the hot pool coolant. However, the reactor vessel wall temperature is still calculated by a simple lumped model

Behavior of LWR fuel elements under accident conditions

International Nuclear Information System (INIS)

Albrecht, H.; Bocek, M.; Erbacher, F.; Fiege, A.; Fischer, M.; Hagen, S.; Hofmann, P.; Holleck, H.; Karb, E.; Leistikow, S.; Melang, S.; Ondracek, G.; Thuemmler, F.; Wiehr, K.

1977-01-01

In the frame of the German reactor safety research program, the Kernforschungszentrum Karlsruhe is carrying out a comprehensive program on the behavior of LWR fuel elements under a variety of power cooling mismatch conditions in particular during loss-of-coolant accidents. The major objectives are to establish a detailed quantitative understanding of fuel rod failures mechanisms and their thresholds, to evaluate the safety margins of power reactor cores under accident conditions and to investigate the feedback of fuel rod failures on the efficiency of emergency core cooling systems. This detailed quantitative understanding is achieved through extensive basic and integral experiments and is incorporated in a fuel behavior code. On the basis of these results the design of power reactor fuel elements and of safety devices can be further improved. The results of investigations on the inelastic deformation (ballooning) behavior of Zircaloy 4 cladding at LOCA temperatures in oxidizing atmosphere are presented. Depending upon strain rate and temperature superplastic deformation behavior was observed. In the equation of state of Zry 4 the strain rate sensitivity index depends strongly upon strain and in the superplastic region upon sample anisotropy. Oxidation kinetics experiments with Zry-tubes at 900-1300 0 C showed that the Baker-Just correlation describes the reality quite conservative. Therefore a reduction of the amount of Zry oxidation can be assumed in the course of a LOCA. The external oxidation of Zry-cladding by steam as well as internal oxidation by the oxygen in oxide fuel and fission products (Cs, I, Te) have an influence on the strain and rupture behavior of Zry-cladding at LOCA temperatures. In out-of-pile and inpile experiments the mechanical and thermal behavior of fuel rods during the blowdown, the heatup and the reflood phases of a LOCA are investigated under representative and controlled thermohydraulic conditions. The task of the inpile experiments is

Radiologic inspection in an office built rod contaminated with radioactive material, in Tiaquepaque, Jalisco, Mexico; Inspeccion radiologica en una oficina construida con varilla contaminada con material radiactivo, en Tlaquepaque, Jalisco, Mexico

Energy Technology Data Exchange (ETDEWEB)

Salas Mar, B.

2011-07-01

Note that in 1983 took place {sup T}he most important radiological accident occurred in Mexico in recent years occurred in Ciudad Juarez, Chihuahua, consisting ea involuntary casting a source of Cobalt-60 (originally intended for medical use in teletherapy) and manufacture of rods for the construction and steel bases for tables, in the smelter {sup S}teel de Chihuahua{sup ,} so presumably, that the rods of the office of Tlaquepaque, come from the radiological accident.

BUSH: A computer code for calculating steady state heat transfer in LWR rod bundles under accident conditions

International Nuclear Information System (INIS)

Shepherd, I.M.

1982-01-01

The computer code BUSH has been developed for the calculation of steady state heat transfer in a rod bundle. For a given power, flow and geometry it can calculate the temperatures in the rods, coolant and shroud assuming that at any axial level each rod can be described by one temperature and the coolant fluid is also radially uniform at this level. Heat transfer by convection and radiation are handled and the geometry is flexible enough to model nearly all types of envisaged shroud design for the SUPERSARA test series. The modular way in which BUSH has been written makes it suitable for future development, either within the present BUSH framework or as part of a more advanced code

Proton ejection project for Saturne

International Nuclear Information System (INIS)

Bronca, G.; Gendreau, G.

1959-01-01

The reasons for choosing the ejection system are given. The characteristics required for the ejected beam are followed by a description of the ejection process, in chronological order from the viewpoint of the protons: movement of the particles, taking into account the various elements which make up the system (internal magnet, external magnet, quadrupoles, ejection correction coils, thin and thick cables,...) and specification of these elements. Then follows an estimation of the delay in manufacture and the cost of the project. Finally, the characteristics of the magnets and quadrupoles are listed in an appendix. (author) [fr

Prediction of failure enthalpy and reliability of irradiated fuel rod under reactivity-initiated accidents by means of statistical approach

International Nuclear Information System (INIS)

Nam, Cheol; Choi, Byeong Kwon; Jeong, Yong Hwan; Jung, Youn Ho

2001-01-01

During the last decade, the failure behavior of high-burnup fuel rods under RIA has been an extensive concern since observations of fuel rod failures at low enthalpy. Of great importance is placed on failure prediction of fuel rod in the point of licensing criteria and safety in extending burnup achievement. To address the issue, a statistics-based methodology is introduced to predict failure probability of irradiated fuel rods. Based on RIA simulation results in literature, a failure enthalpy correlation for irradiated fuel rod is constructed as a function of oxide thickness, fuel burnup, and pulse width. From the failure enthalpy correlation, a single damage parameter, equivalent enthalpy, is defined to reflect the effects of the three primary factors as well as peak fuel enthalpy. Moreover, the failure distribution function with equivalent enthalpy is derived, applying a two-parameter Weibull statistical model. Using these equations, the sensitivity analysis is carried out to estimate the effects of burnup, corrosion, peak fuel enthalpy, pulse width and cladding materials used

A survey of blockage measurement methods used in PWR multi-rod experiments

Energy Technology Data Exchange (ETDEWEB)

Hindle, E.D.; Jones, C.; Whitty, S. (AEA Reactor Services, Springfield (UK))

1986-05-01

The deformation characteristics of Zircaloy multi-rod arrays are being investigated in laboratory and in-reactor tests, and heat transfer experiments are being carried out on pre-deformed arrays. The primary objective is to demonstrate that cladding distension occurring under hypothetical loss-of-coolant accident (LOCA) conditions will not impede the PWR emergency coolant flow during the reflood stage to the extent that unacceptably high cladding temperatures are reached, i.e. that a coolable geometry is maintained. This Report critically reviews the current methods for measuring blockage in multi-rod arrays and discusses their application. A new definition which overcomes the deficiencies of the previous methods is proposed even though it still has drawbacks in the case of overall blockage measurement. A method for automatically measuring the individual rod strain, general cluster blockage sub-channel blockage and sub-channel perimeter changes is described and the results from a deformed array presented. (author).

A survey of blockage measurement methods used in PWR multi-rod experiments

International Nuclear Information System (INIS)

Hindle, E.D.; Jones, C.; Whitty, S.

1986-05-01

The deformation characteristics of Zircaloy multi-rod arrays are being investigated in laboratory and in-reactor tests, and heat transfer experiments are being carried out on pre-deformed arrays. The primary objective is to demonstrate that cladding distension occurring under hypothetical loss-of-coolant accident (LOCA) conditions will not impede the PWR emergency coolant flow during the reflood stage to the extent that unacceptably high cladding temperatures are reached, i.e. that a coolable geometry is maintained. This Report critically reviews the current methods for measuring blockage in multi-rod arrays and discusses their application. A new definition which overcomes the deficiencies of the previous methods is proposed even though it still has drawbacks in the case of overall blockage measurement. A method for automatically measuring the individual rod strain, general cluster blockage sub-channel blockage and sub-channel perimeter changes is described and the results from a deformed array presented. (author)

A user input manual for single fuel rod behaviour analysis code FEMAXI-III

International Nuclear Information System (INIS)

Saito, Hiroaki; Yanagisawa, Kazuaki; Fujita, Misao.

1983-03-01

Principal objectives of Safety related research in connection with lighr water reactor fuel rods under normal operating condition are mainly addressed 1) to assess fuel integrity under steady state condition and 2) to generate initial condition under hypothetical accident. These assessments have to be relied principally upon steady state fuel behaviour computing code that is able to calculate fuel conditions to tbe occurred in a various manner. To achieve these objectives, efforts have been made to develope analytical computer code that calculates in-reactor fuel rod behaviour in best estimate manner. The computer code developed for the prediction of the long-term burnup response of single fuel rod under light water reactor condition is the third in a series of code versions:FEMAMI-III. The code calculates temperature, rod internal gas pressure, fission gas release and pellet-cladding interaction related rod deformation as a function of time-dependent fuel rod power and coolant boundary conditions. This document serves as a user input manual for the code FEMAMI-III which has opened to the public in year of 1982. A general description of the code input and output are included together with typical examples of input data. A detailed description of structures, analytical submodels and solution schemes in the code shall be given in the separate document to be published. (author)

Reactivity initiated accident test series Test RIA 1-4 fuel behavior report

International Nuclear Information System (INIS)

Cook, B.A.; Martinson, Z.R.

1984-09-01

This report presents and discusses results from the final test in the Reactivity Initiated Accident (RIA) Test Series, Test RIA 1-4, conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. Nine preirradiated fuel rods in a 3 x 3 bundle configuration were subjected to a power burst while at boiling water reactor hot-startup system conditions. The test resulted in estimated axial peak, radial average fuel enthalpies of 234 cal/g UO 2 on the center rod, 255 cal/g UO 2 on the side rods, and 277 cal/g UO 2 on the corner rods. Test RIA 1-4 was conducted to investigate fuel coolability and channel blockage within a bundle of preirradiated rods near the present enthalpy limit of 280 cal/g UO 2 established by the US Nuclear Regulatory Commission. The test design and conduct are described, and the bundle and individual rod thermal and mechanical responses are evaluated. Conclusions from this final test and the entire PBF RIA Test Series are presented

Control rod drives

International Nuclear Information System (INIS)

Nakamura, Akira.

1984-01-01

Purpose: To enable to monitor the coupling state between a control rod and a control rod drive. Constitution: After the completion of a control rod withdrawal, a coolant pressure is applied to a control rod drive being adjusted so as to raise only the control rod drive and, in a case where the coupling between the control rod drive and the control rod is detached, the former is elevated till it contacts the control rod and then stopped. The actual stopping position is detected by an actual position detection circuit and compared with a predetermined position stored in a predetermined position detection circuit. If both of the positions are not aligned with each other, it is judged by a judging circuit that the control rod and the control rod drives are not combined. (Sekiya, K.)

Radioactive release during nuclear accidents in Chernobyl and Fukushima

Science.gov (United States)

Nur Ain Sulaiman, Siti; Mohamed, Faizal; Rahim, Ahmad Nabil Ab

2018-01-01

Nuclear accidents that occurred in Chernobyl and Fukushima have initiated many research interests to understand the cause and mechanism of radioactive release within reactor compound and to the environment. Common types of radionuclide release are the fission products from the irradiated fuel rod itself. In case of nuclear accident, the focus of monitoring will be mostly on the release of noble gases, I-131 and Cs-137. As these are the only accidents have been rated within International Nuclear Events Scale (INES) Level 7, the radioactive release to the environment was one of the critical insights to be monitored. It was estimated that the release of radioactive material to the atmosphere due to Fukushima accident was approximately 10% of the Chernobyl accident. By referring to the previous reports using computational code systems to model the release rate, the release activity of I-131 and Cs-137 in Chernobyl was significantly higher compare to Fukushima. The simulation code also showed that Chernobyl had higher release rate of both radionuclides on the day of accident. Other factors affecting the radioactive release for Fukushima and Chernobyl accidents such as the current reactor technology and safety measures are also compared for discussion.

Control rod assembly

International Nuclear Information System (INIS)

Takahashi, Akio.

1982-01-01

Purpose: To enable reliable insertion and drops of control rods, as well as insure a sufficient flow rate of coolants flowing through the control rods for attaining satisfactory cooling thereof to enable relexation of thermal stress resulted to rectifying mechanisms or the likes. Constitution: To the outer circumference of a control rod contained vertically movably within a control rod guide tube, resistive members are retractably provided in such a way as to project to close the gap between outer circumference of the control rod and the inner surface of the control rod guide tube upon engagement of a gripper of control rod drives, and retract upon release of the engagement of the gripper. Thus, since the resistive members project to provide a greater resistance to the coolants flowing between them and the control rod guide tube in the normal operation where the gripper is engaged to drive the control rod by the control rod drives, a major part of the coolant flowing into the control rod guide tube flows into the control rod. This enables to cool the control rod effectively and make the temperature distribution uniform for the coolant flowing from the upper end of the control rod guide tube to thereby attain the relaxation of the thermal stress resulted in the rectifying mechanisms or the likes. (Moriyama, K.)

Device for coupling a control rod and control rod drive

International Nuclear Information System (INIS)

Nishioka, Kazuya.

1975-01-01

Object: To obtain simple and reliable coupling between a control rod and control rod drive by equipping the lower end of the control rod with an extension provided with lateral protuberances and forming the upper end of an index tube with a recess provided with lateral holes. Structure: The tapering central extension of the control rod is inserted into the recess by lowering the control rod, and then it is further inserted by causing frictional movement of the inclined surfaces of lateral protuberances in frictional contact with guide surfaces. When the lateral protuberances are brought into contact with a stepped portion, the control rod is rotated to fit the lateral protuberances into the lateral holes. In this way, the control rod is coupled to the index tube of the control rod drive. (Yoshino, Y.)

A parametric thermohydraulic study an advanced pressurized light water reactor with a tight fuel rod lattice

International Nuclear Information System (INIS)

Dalle Donne, M.; Hame, W.

1982-12-01

A parametric thermohydraulic study for an Advanced Pressurized Light Water Reactor (APWR) with a tight fuel rod lattice has been performed. The APWR improves the uranium utilisation. The APWR core should be placed in a modern German PWR plant. Within this study about 200 different reactors have been calculated. The tightening of the fuel rod lattice implies a decrease of the net electrical output of the plant, which is greater for the heterogeneous reactor than for the homogeneous reactor. APWR cores mean higher core pressure drops and higher water velocities in the core region. The cores tend to be shorter and the number of fuel rods to be higher than for the PWR. At the higher fuel rod pitch to diameter ratios (p/d) the DNB limitation is more stringent than the limitation on the fuel rod linear rating given by the necessity of reflooding after a reactor accident. The contrary is true for the lower p/d ratios. Subcooled boiling in the highest rated coolant channels occurs for the most of the calculated reactors. (orig.) [de

ORNL rod-bundle heat-transfer test data. Volume 7. Thermal-Hydraulic Test Facility experimental data report for test series 3.07.9 - steady-state film boiling in upflow

International Nuclear Information System (INIS)

Mullins, C.B.; Felde, D.K.; Sutton, A.G.; Gould, S.S.; Morris, D.G.; Robinson, J.J.

1982-05-01

Thermal-Hydraulic Test Facility (THTF) test series 3.07.9 was conducted by members of the Oak Ridge National Laboratory Pressurized-Water Reactor (ORNL-PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on September 11, September 18, and October 1, 1980. The objective of the program is to investigate heat transfer phenomena believed to occur in PWRs during accidents, including small- and large-break loss-of-coolant accidents. Test series 3.07.9 was designed to provide steady-state film boiling data in rod bundle geometry under reactor accident-type conditions. This report presents the reduced instrument responses for THTF test series 3.07.9. Also included are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers

PCDP [Prototypical Spent Fuel Consolidation Equipment Demonstration Project] design basis accident report 9315-P-103, Rev. A

International Nuclear Information System (INIS)

1987-12-01

The Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM) has identified a requirement to integrate the spent fuel rod consolidation design activities of each of several proposed geological repository facilities and the Monitored Retrievable Storage (MRS) facility, and to develop efficient and cost-effective equipment for the consolidation process. The equipment to be developed for the rod consolidation system will be required to operate in a dry environment at rates which can be appropriately scaled to approximate the waste management system acceptance rates, irrespective of repository geologic characteristics or the existence of an MRS facility in the waste management system. The purpose of this report is to identify and analyze the range of facility credible events and accident occurrences (from minor to the design basis accidents) and their causes and consequences. For each situation, the considerations to prevent or mitigate the event or accident is addressed

TITAN: an advanced three-dimensional coupled neutronic/thermal-hydraulics code for light water nuclear reactor core analysis

International Nuclear Information System (INIS)

Griggs, D.P.; Kazimi, M.S.; Henry, A.F.

1984-06-01

The three-dimensional nodal neutronics code QUANDRY and the three-dimensional two-fluid thermal-hydraulics code THERMIT are combined into TITAN. Steady-state and transient coupling methodologies based upon a tandem structure were devised and implemented. Additional models for nuclear feedback, equilibrium xenon and direct moderator heating were added. TITAN was tested using a boiling water two channel problem and the coupling methodologies were shown to be effective. Simulated turbine trip transients and several control rod withdrawal transients were analyzed with good results. Sensitivity studies indicated that the time-step size can affect transient results significantly. TITAN was also applied to a quarter core PWR problem based on a real reactor geometry. The steady-state results were compared to a solution produced by MEKIN-B and poor agreement between the horizontal power shapes was found. Calculations with various mesh spacings showed that the mesh spacings in the MEKIN-B analysis were too large to produce accurate results with a finite difference method. The TITAN results were shown to be reasonable. A pair of control rod ejection accidents were also analyzed with TITAN. A comparison of the TITAN PWR control rod ejection results with results from coupled point kinetics/thermal-hydraulics analyses showed that the point kinetics method used (adiabatic method for control rod reactivities, steady-state flux shape for core-averaged reactivity feedback) underpredicted the power excursion in one case and overpredicted it in the other. It was therefore concluded that point kinetics methods should be used with caution and that three-dimensional codes like TITAN are superior for analyzing PWR control rod ejection transients

Control rod displacement

International Nuclear Information System (INIS)

Nakazato, S.

1987-01-01

This patent describes a nuclear reactor including a core, cylindrical control rods, a single support means supporting the control rods from their upper ends in spaced apart positions and movable for displacing the control rods in their longitudinal direction between a first end position in which the control rods are fully inserted into the core and a second end position in which the control rods are retracted from the core, and guide means contacting discrete regions of the outer surface of each control rod at least when the control rods are in the vicinity of the second end position. The control rods are supported by the support means for longitudinal movement without rotation into and out of the core relative to the guide means to thereby cause the outer surface of the control rods to experience wear as a result of sliding contact with the guide means. The support means are so arranged with respect to the core and the guide means that it is incapable of rotation relative to the guide means. The improvement comprises displacement means being operatively coupled to a respective one of the control rods for periodically rotating the control rod in a single angular direction through an angle selected to change the locations on the outer surfaces of the control rods at which the control rods are contacted by the guide means during subsequent longitudinal movement of the control rods

Calculation of Equivalent Resistance for Ground Wires Twined with Armor Rods in Contact Terminals

Directory of Open Access Journals (Sweden)

Gang Liu

2018-03-01

Full Text Available Ground wire breakage accidents can destroy the stable operation of overhead lines. The excessive temperature increase arising from the contact resistance between the ground wire and armor rod in the contact terminal is one of the main reasons causing the breakage of ground wires. Therefore, it is necessary to calculate the equivalent resistance for ground wires twined with armor rods in contact terminals. According to the actual distribution characteristics of the contact points in the contact terminal, a three-dimensional electromagnetic field simulation model of the contact terminal was established. Based on the model, the current distribution in the contact terminal was obtained. Subsequently, the equivalent resistance of a ground wire twined with the armor rod in the contact terminal was calculated. The effects of the factors influencing the equivalent resistance were also discussed. The corresponding verification experiments were conducted on a real ground wire on a contact terminal. The measurement results of the equivalent resistance for the armor rod segment showed good agreement with the electromagnetic modeling results.

Morphoelastic rods. Part I: A single growing elastic rod

KAUST Repository

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.

Morphoelastic rods. Part I: A single growing elastic rod

KAUST Repository

Moulton, D.E.; Lessinnes, T.; Goriely, A.

2013-01-01

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.

Experiments with preirradiated fuel rods in the Nuclear Safety Research Reactor

International Nuclear Information System (INIS)

Horiki, O.; Kobayashi, S.; Takariko, I.; Ishijima, K.

1992-01-01

In the Nuclear Safety Research Reactor (NSRR) owned and operated by Japan Atomic Energy Research Institute (JAERI), extensive experimental studies on the fuel behavior under reactivity initiated accident (RIA) conditions have been continued since the start of the test program in 1975. Accumulated experimental data were used as the fundamental data base of the Japanese safety evaluation guideline for reactivity initiated events in light water cooled nuclear power plants established by the nuclear safety commission in 1984. All of the data used to establish the guideline were, however, limited to those derived from the tests with fresh fuel rods as test samples because of the lack of experimental facility to handle highly radioactive materials.The guideline, therefore, introduces the peak fuel enthalpy of 85 cal/g which was adopted from the SPERT-CDC data as a provisional failure threshold of preirradiated fuel rod and, says that this value should be revised based on the NSRR experiments in the future. According to the above requirement, new NSRR experimental program with the preirradiated fuel rods as test samples was started in 1989. Test fuel rods are prepared by refabrication of the long-sized fuel rods preirradiated in commercial PWRs and BWRs into short segments and by preirradiation of short-sized test fuel rods in the Japan Material Testing Reactor(JMTR). For the tests with preirradiated fuel rods as test samples, the special experimental capsules, the automatic instrumentation fitting device, the automatic capsule assembling device and the capsule loading device were newly developed. In addition, the existing hot cave was modified to mount the capsule assembling device and the other inspection tools and, a new small iron cell was established adjacent to the cave to store the instrumentation fitting device. (author)

Control rod drive

International Nuclear Information System (INIS)

Hawke, B.C.

1986-01-01

A reactor core, one or more control rods, and a control rod drive are described for selectively inserting and withdrawing the one or more control rods into and from the reactor core, which consists of: a support structure secured beneath the reactor core; control rod positioning means supported by the support structure for movably supporting the control rod for movement between a lower position wherein the control rod is located substantially beneath the reactor core and an upper position wherein at least an upper portion of the control rod extends into the reactor core; transmission means; primary drive means connected with the control rod positioning means by the transmission means for positioning the control rod under normal operating conditions; emergency drive means for moving the control rod from the lower position to the upper position under emergency conditions, the emergency drive means including a weight movable between an upper and a lower position, means for movably supporting the weight, and means for transmitting gravitational force exerted on the weight to the control rod positioning means to move the control rod upwardly when the weight is pulled downwardly by gravity; the transmission means connecting the control rod positioning means with the emergency drive means so that the primary drive means effects movement of the weight and the control rod in opposite directions under normal conditions, thus providing counterbalancing to reduce the force required for upward movement of the control rod under normal conditions; and restraint means for restraining the fall of the weight under normal operating conditions and disengaging the primary drive means to release the weight under emergency conditions

Higher-speed coronal mass ejections and their geoeffectiveness

Science.gov (United States)

Singh, A. K.; Bhargawa, Asheesh; Tonk, Apeksha

2018-06-01

We have attempted to examine the ability of coronal mass ejections to cause geoeffectiveness. To that end, we have investigated total 571 cases of higher-speed (> 1000 km/s) coronal mass ejection events observed during the years 1996-2012. On the basis of angular width (W) of observance, events of coronal mass ejection were further classified as front-side or halo coronal mass ejections (W = 360°); back-side halo coronal mass ejections (W = 360°); partial halo (120°mass ejections were much faster and more geoeffective in comparison of partial halo and non-halo coronal mass ejections. We also inferred that the front-sided halo coronal mass ejections were 67.1% geoeffective while geoeffectiveness of partial halo coronal mass ejections and non-halo coronal mass ejections were found to be 44.2% and 56.6% respectively. During the same period of observation, 43% of back-sided CMEs showed geoeffectiveness. We have also investigated some events of coronal mass ejections having speed > 2500 km/s as a case study. We have concluded that mere speed of coronal mass ejection and their association with solar flares or solar activity were not mere criterion for producing geoeffectiveness but angular width of coronal mass ejections and their originating position also played a key role.

ORNL rod-bundle heat-transfer test data. Volume 3. Thermal-hydraulic test facility experimental data report for test 3.06.6B - transient film boiling in upflow

International Nuclear Information System (INIS)

Mullins, C.B.; Felde, D.K.; Sutton, A.G.; Gould, S.S.; Morris, D.G.; Robinson, J.J.

1982-05-01

Reduced instrument responses are presented for Thermal-Hyraulic Test Facility (THTF) Test 3.06.6B. This test was conducted by members of the Oak Ridge National Laboratory Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on August 29, 1980. The objective of the program was to investigate heat transfer phenomena believed to occur in PWR's during accidents, including small and large break loss-of-coolant accidents. Test 3.06.6B was conducted to obtain transient film boiling data in rod bundle geometry under reactor accident-type conditions. The primary purpose of this report is to make the reduced instrument responses for THTF Test 3.06.6B available. Included in the report are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers

The time-dependent 3D discrete ordinates code TORT-TD with thermal-hydraulic feedback by ATHLET models

International Nuclear Information System (INIS)

Seubert, A.; Velkov, K.; Langenbuch, S.

2008-01-01

This paper describes the time-dependent 3D discrete ordinates transport code TORT-TD. Thermal-hydraulic feedback is considered by coupling TORT-TD with the thermal-hydraulics system code ATHLET. The coupled code TORT-TD/ATHLET allows 3D pin-by-pin analyses of transients in few energy groups and anisotropic scattering by solving the time-dependent transport equation using the unconditionally stable implicit method. The nuclear cross sections are interpolated between pre-calculated table values of fuel temperature, moderator density and boron concentration. For verification of the implementation, selected test cases have been calculated by TORT-TD/ATHLET. They include a control rod ejection transient in a small PWR fuel assembly arrangement and a local boron concentration change in a single PWR fuel assembly. In the latter, special attention has been paid to study the influence of the thermal-hydraulic feedback modelling in ATHLET. The results obtained for a control rod ejection accident in a PWR quarter core demonstrate the applicability of TORT-TD/ATHLET. (authors)

Analysis on the nitrogen drilling accident of Well Qionglai 1 (I: Major inducement events of the accident

Directory of Open Access Journals (Sweden)

Yingfeng Meng

2015-12-01

Full Text Available Nitrogen drilling in poor tight gas sandstone should be safe because of very low gas production. But a serious accident of fire blowout occurred during nitrogen drilling of Well Qionglai 1. This is the first nitrogen drilling accident in China, which was beyond people's knowledge about the safety of nitrogen drilling and brought negative effects on the development of gas drilling technology still in start-up phase and resulted in dramatic reduction in application of gas drilling. In order to form a correct understanding, the accident was systematically analyzed, the major events resulting in this accident were inferred. It is discovered for the first time that violent ejection of rock clasts and natural gas occurred due to the sudden burst of downhole rock when the fractured tight gas zone was penetrated during nitrogen drilling, which has been named as “rock burst and blowout by gas bomb”, short for “rock burst”. Then all the induced events related to the rock burst are as following: upthrust force on drilling string from rock burst, bridging-off formed and destructed repeatedly at bit and centralizer, and so on. However, the most direct important event of the accident turns out to be the blockage in the blooie pipe from rock burst clasts and the resulted high pressure at the wellhead. The high pressure at the wellhead causes the blooie pipe to crack and trigged blowout and deflagration of natural gas, which is the direct presentation of the accident.

A thermohydraulic analysis for LOCA accident of a CANDU 600 reactor core charged with SEU 43 fuel by means of FIREBIRD code

International Nuclear Information System (INIS)

Serbanel, M.; Catana, A.

2001-01-01

This report presents a comparative analysis of the behaviour of primary circuit during a LOCA 20% RIH accident for two types of reactor core, namely, normally charged, i.e., with clusters of 37 rods and charged with clusters of 43 rods, respectively. This type of accident was chosen since Canadian analyses showed that the associated transient regime stress the fuel elements. The void reactivity as a function of coolant average density was calibrated for a reference regime (LOCA 20% RIH) so that the results of the model be able to reproduce the average distribution in the reference transient regime. The computation makes use of CERBERUS and FIREBIRD codes externally coupled by files. The void reactivity of the hot pencil was obtained this way. An extremely conservative hypothesis was used, namely that the momentary power of the cluster hosting the pencil is the maximal power over the cluster for the corresponding half reactor core. To carry out this work the following steps were covered: 1. The scenario for the LOCA 20% RIH accident was worked out and the input data corresponding to the thermohydraulic and neutronic modules, for the complex model and the 37 rod clusters, were checked; 2. The input data corresponding to the thermohydraulic module for the complex model and the 43 rod cluster were checked; 3. The kinetic parameters corresponding to the 37 rod cluster were computed; 4. The kinetic parameters corresponding to the 43 rod cluster were computed and the file for the input data in the neutronic module was built; 5. A sub-routine for writing files with the thermohydraulic and neutronic quantities, in a format adequate to the other programs, was implemented; 6. The two transient regimes considered were implemented and the archives containing the quantities were built ;7. The results obtained were analyzed. The conclusion of this work is that in case of LOCA 20% RIH accident the 43 bar clusters have a better behaviour than the 37 bar clusters

Experiment data report for Test RIA 1-2 (Reactivity Initiated Accident Test Series)

International Nuclear Information System (INIS)

Zimmermann, C.L.; White, C.E.; Evans, R.P.

1979-06-01

Recorded test data are presented for the second of six planned tests in the Reactivity Initiated Accident (RIA) Test Series I, Test RIA 1-2. This test, conducted at the Power Burst Facility, had the following objectives: (1) characterize the response of preirradiated fuel rods during an RIA event conducted at boiling water reactor hot-startup conditions; and (2) evaluate the effect of rod internal pressure on preirradiated fuel rod response during an RIA event. The data from Test RIA 1-2 are graphed in engineering units and have been appraised for quality and validity. These uninterpreted data are presented for use in the nuclear fuel behavior research field before detailed analysis and interpretation have been completed

Analysis of eventual accidents in a water experimental loop, using the Relap 4 computer code

International Nuclear Information System (INIS)

Fernandes Filho, T.L.

1981-01-01

Transients caused by accidents as (1) loss of coolant, (2) failure in the principal pump and (3) power excursions were analysed. In the accident simulation, the Relap 4/Mod 3 computer code was used. The results obtained with the steady state model showed to be consistent with the project-and operation data of the experimental loop. For all the accidents analysed that considered the performance of safety systems, the highest temperature of the heating rods in the testing section did not exceed the permissible temperature. (E.G.) [pt

Dealing with Historical Discrepancies: The Recovery of National Research Experiment (NRX) Reactor Fuel Rods at Chalk River Laboratories (CRL) - 13324

International Nuclear Information System (INIS)

Vickerd, Meggan

2013-01-01

Following the 1952 National Research Experiment (NRX) Reactor accident, fuel rods which had short irradiation histories were 'temporarily' buried in wooden boxes at the 'disposal grounds' during the cleanup effort. The Nuclear Legacy Liabilities Program (NLLP), funded by Natural Resources Canada (NRCan), strategically retrieves legacy waste and restores lands affected by Atomic Energy of Canada Limited (AECL) early operations. Thus under this program the recovery of still buried NRX reactor fuel rods and their relocation to modern fuel storage was identified as a priority. A suspect inventory of NRX fuels was compiled from historical records and various research activities. Site characterization in 2005 verified the physical location of the fuel rods and determined the wooden boxes they were buried in had degraded such that the fuel rods were in direct contact with the soil. The fuel rods were recovered and transferred to a modern fuel storage facility in 2007. Recovered identification tags and measured radiation fields were used to identify the inventory of these fuels. During the retrieval activity, a discrepancy was discovered between the anticipated number of fuel rods and the number found during the retrieval. A total of 32 fuel rods and cans of cut end pieces were recovered from the specified site, which was greater than the anticipated 19 fuel rods and cans. This discovery delayed the completion of the project, increased the associated costs, and required more than anticipated storage space in the modern fuel storage facility. A number of lessons learned were identified following completion of this project, the most significant of which was the potential for discrepancies within the historical records. Historical discrepancies are more likely to be resolved by comprehensive historical record searches and site characterizations. It was also recommended that a complete review of the wastes generated, and the total affected lands as a result of this historic

OECD/DOE/CEA VVER-1000 coolant transient (V1000CT) benchmark - a consistent approach for assessing coupled codes for RIA analysis

International Nuclear Information System (INIS)

Boyan D Ivanov; Kostadin N Ivanov; Eric Royer; Sylvie Aniel; Nikola Kolev; Pavlin Groudev

2005-01-01

Full text of publication follows: The Rod Ejection Accident (REA) and Main Steam Line Break (MSLB) are two of the most important Design Basis Accidents (DBA) for VVER-1000 exhibiting significant localized space-time effects. A consistent approach for assessing coupled three-dimensional (3-D) neutron kinetics/thermal hydraulics codes for these Reactivity Insertion Accidents (RIA) is to first validate the codes using the available plant test (measured) data and after that perform cross code comparative analysis for REA and MSLB scenarios. In the framework of joint effort between the Nuclear Energy Agency (NEA) of OECD, the United States Department of Energy (US DOE), and the Commissariat a l'Energie Atomique (CEA), France a coupled 3-D neutron kinetics/thermal hydraulics benchmark was defined. The benchmark is based on data from the Unit 6 of the Bulgarian Kozloduy Nuclear Power Plant (NPP). In performing this work the PSU, USA and CEA-Saclay, France have collaborated with Bulgarian organizations, in particular with the KNPP and the INRNE. The benchmark consists of two phases: Phase 1: Main Coolant Pump Switching On; Phase 2: Coolant Mixing Tests and MSLB. In addition to the measured (experiment) scenario, an extreme calculation scenario was defined for better testing 3-D neutronics/thermal-hydraulics techniques: rod ejection simulation with control rod being ejected in the core sector cooled by the switched on MCP. Since the previous coupled code benchmarks indicated that further development of the mixing computation models in the integrated codes is necessary, a coolant mixing experiment and MSLB transients are selected for simulation in Phase 2 of the benchmark. The MSLB event is characterized by a large asymmetric cooling of the core, stuck rods and a large primary coolant flow variation. Two scenarios are defined in Phase 2: the first scenario is taken from the current licensing practice and the second one is derived from the original one using aggravating

Analytical and experimental assessment of TVS-2006 fuel assembly thermal-mechanical shape deformation at temperature modeling of a loss-of-coolant accident

International Nuclear Information System (INIS)

Afanasiev, A.; Semishkin, V.; Makarov, V.; Matvienko, I.; Puzanov, D.

2015-01-01

Full or partial core drying-out takes place in loss-of-coolant accidents, which leads to worsening of heat removal from the fuel rods. Depending on the accident scenario the fuel rod cladding temperature can be in a wide range from 350 to 1200°C. It is worth mentioning, that the length of the process can considerably affect the fuel rod cladding loadcarrying capacity and the FA structure as a whole, and in the long run it defines the radiation consequences of the accident and the possibility of postaccident core disassembly at low cost. Most experiments staged of late were devoted to a study of FA behaviour in the temperature range 800-900°C of α→β phase transition that is characterized by a sharp increase in the rate of zirconium alloy creep which leads to fuel rod cladding ballooning and loss of their tightness within a short period of time. The 600-700°C temperature range turned out to be less investigated whereas this is the range where the change of zirconium alloy mechanical properties is also observed but only with the retention of α-phase. The tests of a full-scale FA dummy with the skeleton of guide tubes and spacer grids connected by friction forces, carried out at the testing facility of JSC OKB “GIDROPRESS”, were devoted to a study of FA behaviour in this temperature range. The model was heated up with hot air to 650°C for 6 hours. The tests ended with fuel rod cladding ballooning due to gauge pressure and shape deformation. No loss of fuel rod cladding integrity was observed. Therefore, a conclusion can be made that a long-time core holdup at the parameters implemented at the test facility is permitted and the deformations of the FA structure do not lead to the damage that could considerably complicate the core disassembly. The test results were used for the verification of the calculational model of FA TVS-2006 structure with a welded skeleton by ANSYS code. On the basis of the verified calculational model a calculational model was

Aircrew ejection experience: questionnaire responses from 20 survivors.

Science.gov (United States)

Taneja, Narinder; Pinto, Leslie J; Dogra, Manmohan

2005-07-01

Published studies on ejection have focused predominantly on the injuries sustained by aircrew and discussed their preventive measures from an aeromedical perspective. However, studies have not discussed aircrew experiences related to ejection or how they would like to advise other aircrew to successfully handle ejection as an event. Such information can assist in designing realistic indoctrination and training programs. This study was conducted to fill gaps in our understanding of aircrew perspectives of successful ejections. Aircrew reporting to the Institute of Aerospace Medicine (IAM), Indian Air Force, for post-ejection evaluation during the period of May 2003 to January 2005 completed a questionnaire that was designed for the study. A total of 20 aircrew completed this questionnaire. The mean age of the aircrew was 30.25 +/- 4.45 yr. Most of them had logged more than 500 flying hours. Some aircrew described their initial moments of ejection as "blacked out," "dazed, yet conscious," or as "a shock that gradually decreased." Practicing ejection drills on the ground, being prepared at all times, making a timely decision to eject, and assuming correct posture were identified as the most important factors for success. Descriptions of ejection as an event suggest intense emotional arousal could occur following ejection. This study provides first hand inputs into the psychological processes accompanying ejections. Such information could be very useful in understanding the critical factors that influence successful ejection.

Replacement rod

International Nuclear Information System (INIS)

Hatfield, S.C.

1989-01-01

This patent describes in an elongated replacement rod for use with fuel assemblies of the type having two end fittings connected by guide tubes with a plurality of rod and guide tube cell defining spacer grids containing rod support features and mixing vanes. The grids secured to the guide tubes in register between the end fittings at spaced intervals. The fuel rod comprising: an asymmetrically beveled tip; a shank portion having a straight centerline; and a permanently diverging portion between the tip and the shank portion

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

International Nuclear Information System (INIS)

1989-09-01

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)

Steady-state and transient core feasibility analysis for a thorium-fuelled reduced-moderation PWR performing full transuranic recycle

International Nuclear Information System (INIS)

Lindley, Benjamin A.; Ahmad, Ali; Zainuddin, N. Zara; Franceschini, Fausto; Parks, Geoffrey T.

2014-01-01

Highlights: • We present a core analysis for a thorium-transuranic fuelled reduced-moderation PWR. • There is the possibility of positive reactivity in severe large break LOCAs. • Mechanical shim is used to control reactivity within power peaking constraints. • Adequate shutdown margin can be achieved with B 4 C control rods are required. • The response to a rod ejection accident is within likely licensing limits. - Abstract: It is difficult to perform multiple recycle of transuranic (TRU) isotopes in PWRs as the moderator temperature coefficient (MTC) tends to become positive after a few recycles and the core may have positive reactivity when fully voided. Due to the favourable impact on the MTC fostered by use of thorium (Th), the possibility of performing Th–TRU multiple-recycle in reduced-moderation PWRs (RMPWRs) is under consideration. Heterogeneous fuel design with spatial separation of Th–U and Th–TRU is necessary to improve neutronic performance. This can take the form of a heterogeneous fuel assembly (TPUC), or whole assembly heterogeneity (WATU). Satisfactory discharge burn-up can be maintained while ensuring negative MTC, with the pin diameter of a standard PWR increased from 9.5 to 11 mm. However, the reactivity becomes positive when the coolant density in the core becomes extremely low. This could lead to positive reactivity in some loss of coolant accident (LOCA) scenarios, for example a surge line break, if the reactor does not trip. To protect against this beyond design basis accident, a second redundant set of shutdown rods is added to the reactor, so that either the usual or secondary rods can trip the reactor when there is zero coolant in the core. Even so, this condition is likely to be concerning from a regulatory standpoint. Reactivity control is a key challenge due to the reduced worth of neutron absorbers and their detrimental effect on the void coefficients, especially when diluted, as is the case for soluble boron

Water rod

International Nuclear Information System (INIS)

Kashiwai, Shin-ichi; Yokomizo, Osamu; Orii, Akihito.

1992-01-01

In a reactor core of a BWR type reactor, the area of a flow channel in a lower portion of a downcoming pipe for downwardly releasing steams present at the top portion in a water rod is increased. Further, a third coolant flow channel (an inner water rod) is disposed in an uprising having an exit opened near the inlet of the water rod and an inlet opened at the outside near the top portion of the water and having an increase flow channel area in the upper portion. The downcoming pipe in the water rod is filled with steams, and the void ratio is increased by so much as the flow channel area of the downcoming pipe is increased. Since the pressure difference between the inlet and the exit of the inner water rod is greater than the pressure difference between the inlet and the exit of the water rod, most of water flown into the inner water rod is discharged out of the exit in the form of water as it is. Since the area of the flow channel is increased in the portion of the inner water rod, void efficiency in the upper portion of the reactor core is decreased by so much. Since the void ratio is thus increased in the lower portion and the void efficiency is decreased in the upper portion of the reactor core, axial void distribution can be flattened. (N.H.)

Nuclear power plant status diagnostics using artificial neural networks

International Nuclear Information System (INIS)

Bartlett, E.B.; Uhrig, R.E.

1991-01-01

In this work, the nuclear power plant operating status recognition issue is investigated using artificial neural networks (ANNs). The objective is to train an ANN to classify nuclear power plant accident conditions and to assess the potential of future work in the area of plant diagnostics with ANNS. To this end, an ANN was trained to recognize normal operating conditions as well as potentially unsafe conditions based on nuclear power plant training simulator generated accident scenarios. These scenarios include; hot and cold leg loss of coolant, control rod ejection, loss of offsite power, main steam line break, main feedwater line break and steam generator tube leak accidents. Findings show that ANNs can be used to diagnose and classify nuclear power plant conditions with good results

Models of fuel masses transition during second stage of the accident on Chernobyl NPP

International Nuclear Information System (INIS)

Tarapon, A.

2002-01-01

In ISPE NASU of Ukraine are developed mathematical models and software, which allow to research the processes of fuel masses transition during the accident at ChNPP. We found out, that the main reason of accident on ChNPP is the happening in the reactor of crisis of heat exchange of the second sort, instead of the effect positive output of reactivity from displacers of rods of system of emergency protection, as is accepted in official version

Risk impact of two accident management strategies

International Nuclear Information System (INIS)

Dingman, S.; Camp, A.

1992-01-01

This report probabilistic Risk Assessment is used to evaluate two accident management strategies: intentionally depressurizing the reactor coolant system of a pressurized water reactor to prevent containment-pressurization during high pressure melt ejection, and flooding the containment of a boiling water reactor to prevent or delay vessel breach. Sensitivity studies indicated that intentional depressurization would not provide a significant risk reduction at Surry. A preliminary evaluation of the containment flooding strategy indicated that it might prove beneficial for some plants, but that further strategy development would be needed to fully evaluate the strategy-

Control rod

International Nuclear Information System (INIS)

Igarashi, Takao; Yoshimoto, Yuichiro; Sugawara, Satoshi; Fukumoto, Takashi; Endo, Zen-ichiro; Saito, Shozo; Shinpo, Katsutoshi; Nishimura, Akira; Ozawa, Michihiro

1988-01-01

Purpose: To provide a sufficient shutdown margin upon reactor shutdown, prevent sheath deformation without decreasing neutron absorbents and prevent impact shocks exerted to structural materials. Constitution: The control rod of the present invention comprises a neutron absorption region, a sheath deformation means attached to the side wall and means for restricting and supporting axial movement of the neutron absorbent rod. Then, the amount of absorptive nuclei chained absorbents in the lower region is reduced than that in the upper region. In this way, effective neutron absorbing performance can be obtained relative to the neutron importance distribution during reactor shutdown. In addition, since the operationability is improved by reducing the weight of the control rod and the absorptive nuclei chained neutron abosrbers are used, mechanical nuclear life of the control rod can be increased. Thus, it is possible to prevent the outward deformation of the sheath, as well as prevent collision between the neutron absorber rod and the structural material on the side of inserting the control rod generated upon reactor scram by a simple structure. (Kamimura, M.)

Impact of spatial kinetics in severe accident analysis for a large HWR

International Nuclear Information System (INIS)

Morris, E.E.

1994-01-01

The impact on spatial kinetics on the analysis of severe accidents initiated by the unprotected withdrawal of one or more control rods is investigated for a large heavy water reactor. Large inter- and intra-assembly power shifts are observed, and the importance of detailed geometrical modeling of fuel assemblies is demonstrated. Neglect of space-time effects is shown to lead to erroneous estimates of safety margins, and of accident consequences in the event safety margins are exceeded. The results and conclusions are typical of what would be expected for any large, loosely coupled core

Analysis of pressurized water reactor accidents in reactivity disturbances. II

International Nuclear Information System (INIS)

Tinka, I.

1978-01-01

The logic structure of program FATRAP is described. The time course of reactivity temporal and spatial distributions of neutron flux density and power, characteristic temperatures of the individual reactor zones and the heat flux density from cladding to the coolant can be obtained as the main results. The basic program funcitons were tested for a point and a one-dimensional model. In the basic test the absorption rod was removed uncontrollably at a preset speed for 0.5 s with the reactivity feedback operative. A second test simulated the action of the accident protection system with a delay of 0.1 s started when the 7500 MW power had been obtained. The last test consisted in simulating a start-up accident with an initial power of 2.25 MW. For the said chosen accident models reactivity feedback is responsible for the formation of the appropriate power peak while the accident protection attendance alone can considerably reduce temperatures during the process. (J.F.)

Behavior of irradiated ATR/MOX fuel under reactivity initiated accident conditions (Joint research)

International Nuclear Information System (INIS)

Sasajima, Hideo; Fuketa, Toyoshi; Nakamura, Takehiko; Nakamura, Jinichi; Uetsuka, Hiroshi

2000-03-01

Pulse irradiation experiments with irradiated ATR/MOX fuel rods of 20 MWd/kgHM were conducted at the NSRR in JAERI to study the transient behavior of MOX fuel rod under reactivity initiated accident conditions. Four pulse irradiation experiments were performed with peak fuel enthalpy ranging from 335 J/g to 586 J/g, resulted in no failure of fuel rods. Deformation of the fuel rods due to PCMI occurred in the experiments with peak fuel enthalpy above 500 J/g. Significant fission gas release up to 20% was measured by rod puncture measurement. The generation of fine radial cracks in pellet periphery, micro-cracks and boundary separation over the entire region of pellet were observed. These microstructure changes might contribute to the swelling of fuel pellets during the pulse irradiation. This could cause the large radial deformation of fuel rod and high fission gas release when the pulse irradiation conducted at relatively high peak fuel enthalpy. In addition, fine grain structures around the plutonium spot and cauliflower structure in cavity of the plutonium spot were observed in the outer region of the fuel pellet. (author)

Control rod drives

International Nuclear Information System (INIS)

Futatsugi, Masao.

1980-01-01

Purpose: To secure the reactor operation safety by the provision of a fluid pressure detecting section for control rod driving fluid and a control rod interlock at the midway of the flow pass for supplying driving fluid to the control rod drives. Constitution: Between a driving line and a direction control valve are provided a pressure detecting portion, an alarm generating device, and a control rod inhibition interlock. The driving fluid from a driving fluid source is discharged by way of a pump and a manual valve into the reactor in which the control rods and reactor fuels are contained. In addition, when the direction control valve is switched and the control rods are inserted and extracted by the control rod drives, the pressure in the driving line is always detected by the pressure detection section, whereby if abnormal pressure is resulted, the alarm generating device is actuated to warn the abnormality and the control rod inhibition interlock is actuated to lock the direction control valve thereby secure the safety operation of the reactor. (Seki, T.)

Control rod drives

International Nuclear Information System (INIS)

Oonuki, Koji.

1981-01-01

Purpose: To increase the driving speed of control rods at rapid insertion with an elongate control rod and an extension pipe while ensuring sufficient buffering performance in a short buffering distance, by providing a plurality of buffers to an extension pipe between a control rod drive source and a control rod in LMFBR type reactor. Constitution: First, second and third buffers are respectively provided to an acceleration piston, an extension pipe and a control rod respectively and the insertion positions for each of the buffers are displaced orderly from above to below. Upon disconnection of energizing current for an electromagnet, the acceleration piston, the extension pipe and the control rod are rapidly inserted in one body. The first, second and third buffers are respectively actuated at each of their falling strokes upon rapid insertion respectively, and the acceleration piston, the extension pipe and the control rod receive the deceleration effect in the order correspondingly. Although the compression force is applied to the control rod only near the stroke end, it does not cause deformation. (Kawakami, Y.)

TASAC a computer program for thermal analysis of severe accident conditions. Version 3/01, Dec 1991. Model description and user's guide

International Nuclear Information System (INIS)

Stempniewicz, M.; Marks, P.; Salwa, K.

1992-06-01

TASAC (Thermal Analysis of Severe Accident Conditions) is computer code developed in the Institute of Atomic Energy written in FORTRAN 77 for the digital computer analysis of PWR rod bundle behaviour during severe accident conditions. The code has the ability to model an early stage of core degradation including heat transfer inside the rods, convective and radiative heat exchange as well as cladding interactions with coolant and fuel, hydrogen generation, melting, relocations and refreezing of fuel rod materials with dissolution of UO 2 and ZrO 2 in liquid phase. The code was applied for the simulation of International Standard Problem number 28, performed on PHEBUS test facility. This report contains the program physical models description, detailed description of input data requirements and results of code verification. The main directions for future TASAC code development are formulated. (author). 20 refs, 39 figs, 4 tabs

Safety evaluation of accident-tolerant FCM fueled core with SiC-coated zircalloy cladding for design-basis-accidents and beyond DBAs

Energy Technology Data Exchange (ETDEWEB)

Chun, Ji-Han, E-mail: chunjh@kaeri.re.kr; Lim, Sung-Won; Chung, Bub-Dong; Lee, Won-Jae

2015-08-15

Highlights: • Thermal conductivity model of the FCM fuel was developed and adopted in the MARS. • Scoping analysis for candidate FCM FAs was performed to select feasible FA. • Preliminary safety criteria for FCM fuel and SiC/Zr cladding were set up. • Enhanced safety margin and accident tolerance for FCM-SiC/Zr core were demonstrated. - Abstract: The FCM fueled cores proposed as an accident tolerant concept is assessed against the design-basis-accident (DBA) and the beyond-DBA (BDBA) scenarios using MARS code. A thermal conductivity model of FCM fuel is incorporated in the MARS code to take into account the effects of irradiation and temperature that was recently measured by ORNL. Preliminary analyses regarding the initial stored energy and accident tolerant performance were carried out for the scoping of various cladding material candidates. A 16 × 16 FA with SiC-coated Zircalloy cladding was selected as the feasible conceptual design through a preliminary scoping analysis. For a selected design, safety analyses for DBA and BDBA scenarios were performed to demonstrate the accident tolerance of the FCM fueled core. A loss of flow accident (LOFA) scenario was selected for a departure-from-nucleate-boiling (DNB) evaluation, and large-break loss of coolant accident (LBLOCA) scenario for peak cladding temperature (PCT) margin evaluation. A control element assembly (CEA) ejection accident scenario was selected for peak fuel enthalpy and temperature. Moreover, a station blackout (SBO) and LBLOCA without a safety injection (SI) scenario were selected as a BDBA. It was demonstrated that the DBA safety margin of the FCM core is satisfied and the time for operator actions for BDBA s is evaluated.

Power Excursion Accident Analysis of Research Water Reactor

International Nuclear Information System (INIS)

Khaled, S.M.; Doaa, G.M.

2009-01-01

A three-dimensional neutronic code POWEX-K has been developed, and it has been coupled with the sub-channel thermal-hydraulic core analysis code SV based on the Single Mass Velocity Model. This forms the integrated neutronic/thermal hydraulics code system POWEX-K/SV for the accident analysis. The Training and Research Reactors at Budapest University of Technology and Economics (BME-Reactor) has been taken as a reference reactor. The cross-section generation procedure based on WIMS. The code uses an implicit difference approach for both the diffusion equations and thermal-hydraulics modules, with reactivity feedback effects due to coolant and fuel temperatures. The code system was applied to analyzing power excursion accidents initiated by ramp reactivity insertion of 1.2 $. The results show that the reactor is inherently safe in case of such accidents i.e. no core melt is expected even if the safety rods do not fall into the core

Preliminary Design Concept for a Reactor-internal CRDM

International Nuclear Information System (INIS)

Lee, Jae Seon; Kim, Jong Wook; Kim, Tae Wan; Choi, Suhn; Kim, Keung Koo

2013-01-01

A rod ejection accident may cause severer result in SMRs because SMRs have relatively high control rod reactivity worth compared with commercial nuclear reactors. Because this accident would be perfectly excluded by adopting a reactor-internal CRDM (Control Rod Drive Mechanism), many SMRs accept this concept. The first concept was provided by JAERI with the MRX reactor which uses an electric motor with a ball screw driveline. Babcock and Wilcox introduced the concept in an mPower reactor that adopts an electric motor with a roller screw driveline and hydraulic system, and Westinghouse Electric Co. proposes an internal Control Rod Drive in its SMR with an electric motor with a latch mechanism. In addition, several other applications have been reported thus far. The reactor-internal CRDM concept is now widely adopted in many SMR designs, and this concept may also be applied in an evolutionary reactor development. So the preliminary study is conducted based on the SMART CRDM design. A preliminary design concept for a reactor-internal CRDM was proposed and evaluated through an electromagnetic analysis. It was found that there is an optimum design for the motor housing, and the results may contribute to the realization a reactor-internal CRDM for an evolutionary reactor development. More detailed analysis results will be reported later

Control rod cluster with removable rods for nuclear fuel assembly

International Nuclear Information System (INIS)

Denizou, J.P.

1989-01-01

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 [fr

Numerical analysis of the fusion of nuclear combustible rods under LOCA - type accidents

International Nuclear Information System (INIS)

Idelsohn, S.R.; Crivelli, L.A.

1983-01-01

The study of the melting of combustible rods is of great importance for the safety analysis of nuclear reactors. Due to the special characteristics of the problem, a sharp interface between the solid and liquid region does not exist, but appears a 'mushy' region in which the material is partially melted. The Finite Element Method is employed here, together with a regularized enthalpy formulation. Finally, the results obtained are presented and discussed. (Author) [pt

SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays

International Nuclear Information System (INIS)

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

2D model for melt progression through rods and debris

International Nuclear Information System (INIS)

Fichot, F.

2001-01-01

During the degradation of a nuclear core in a severe accident scenario, the high temperatures reached lead to the melting of materials. The formation of liquid mixtures at various elevations is followed by the flow of molten materials through the core. Liquid mixture may flow under several configurations: axial relocation along the rods, horizontal motion over a plane surface such as the core support plate or a blockage of material, 2D relocation through a debris bed, etc.. The two-dimensional relocation of molten material through a porous debris bed, implemented for the simulation of late degradation phases, has opened a new way to the elaboration of the relocation model for the flow of liquid mixture along the rods. It is based on a volume averaging method, where wall friction and capillary effects are taken into account by introducing effective coefficients to characterize the solid matrix (rods, grids, debris, etc.). A local description of the liquid flow is necessary to derive the effective coefficients. Heat transfers are modelled in a similar way. The derivation of the conservation equations for the liquid mixture falling flow (momentum) in two directions (axial and radial-horizontal) and for the heat exchanges (energy) are the main points of this new model for simulating melt progression. In this presentation, the full model for the relocation and solidification of liquid materials through a rod bundle or a debris bed is described. It is implemented in the ICARE/CATHARE code, developed by IPSN in Cadarache. The main improvements and advantages of the new model are: A single formulation for liquid mixture relocation, in 2D, either through a rod bundle or a porous debris bed, Extensions to complex structures (grids, by-pass, etc..), The modeling of relocation of a liquid mixture over plane surfaces. (author)

Safety analysis of RA reactor operation, I-III, Part II, Accident analysis; Analiza sigurnosti rada reaktora RA - I-III, II deo - Analiza akcidenta

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1963-02-15

This volume covers the analyses of two types of accidents: accidents caused by uncontrolled reactivity increase, and accidents caused by decrease or loss of cooling. First type of accidents, uncontrolled reactivity insertion could occur due to removal of compensation, regulatory or safety rods, or by increase of heavy water level. Removal of irradiated samples from the core could also cause increase of reactivity. Second type of accidents could occur due to interruption of cooling, loss of water in the secondary cooling loop or loss of water in the primary coolant loop.

Control rod drive mechanism

International Nuclear Information System (INIS)

Nakamura, Akira.

1981-01-01

Purpose: To ensure the scram operation of a control rod by the reliable detection for the position of control rods. Constitution: A permanent magnet is provided to the lower portion of a connecting rod in engagement with a control rod and a tube having a plurality of lead switches arranged axially therein in a predetermined pitch is disposed outside of the control rod drives. When the control rod moves upwardly in the scram operation, the lead switches are closed successively upon passage of the permanent magnet to operate the electrical circuit provided by way of each of the lead switches. Thus, the position for the control rod during the scram can reliably be determined and the scram characteristic of the control rod can be recognized. (Furukawa, Y.)

TASAC a computer program for thermal analysis of severe accident conditions. Version 3/01, Dec 1991. Model description and user`s guide

Energy Technology Data Exchange (ETDEWEB)

Stempniewicz, M; Marks, P; Salwa, K

1992-06-01

TASAC (Thermal Analysis of Severe Accident Conditions) is computer code developed in the Institute of Atomic Energy written in FORTRAN 77 for the digital computer analysis of PWR rod bundle behaviour during severe accident conditions. The code has the ability to model an early stage of core degradation including heat transfer inside the rods, convective and radiative heat exchange as well as cladding interactions with coolant and fuel, hydrogen generation, melting, relocations and refreezing of fuel rod materials with dissolution of UO{sub 2} and ZrO{sub 2} in liquid phase. The code was applied for the simulation of International Standard Problem number 28, performed on PHEBUS test facility. This report contains the program physical models description, detailed description of input data requirements and results of code verification. The main directions for future TASAC code development are formulated. (author). 20 refs, 39 figs, 4 tabs.

Aspects of severe accidents in transmutation systems

International Nuclear Information System (INIS)

Wider, H.U.; Karlson, J.; Jones, A.V.

2001-01-01

The different types of transmutation systems under investigation include accelerator driven (ADS) and critical systems. To switch off an accelerator in case of an accident initiation is quite important for all accidents. For a fast ADS the grace times available for doing so depend strongly on the total heat capacity and the natural circulation capability of the primary coolant. Cooling with heavy metal Pb-Bi has considerable advantages in this regard compared to gas cooling. Moreover it allows passive ex-vessel cooling with natural air or water circulation. In the remote likelihood of fuel melting, oxide fuel appears to mix with the Pb-Bi coolant. Fast critical systems that are cooled by Pb-Bi will automatically shut off if the flow or heat sink is lost. Reactivity accidents can be limited by a low total control rod worth. High temperature reactors can achieve only incomplete burning of actinides. If an accelerator is added to increase burn-up, a fast spectrum region is needed, which has a low heat capacity. (author)

Method of inserting fuel rod

International Nuclear Information System (INIS)

Kamimoto, Shuji; Imoo, Makoto; Tsuchida, Kenji.

1991-01-01

The present invention concerns a method of inserting a fuel rod upon automatic assembling, automatic dismantling and reassembling of a fuel assembly in a light water moderated reactor, as well as a device and components used therefor. That is, a fuel rod is inserted reliably to an aimed point of insertion by surrounding the periphery of the fuel rod to be inserted with guide rods, and thereby suppressing the movement of the fuel rod during insertion. Alternatively, a fuel rod is inserted reliably to a point of insertion by inserting guide rods at the periphery of the point of insertion for the fuel rod to be inserted thereby surrounding the point of insertion with the guide rods or fuel rods. By utilizing fuel rods already present in the fuel assembly as the guide rods described above, the fuel rod can be inserted reliably to the point of insertion with no additional devices. Dummy fuel rods are previously inserted in a fuel assembly which are then utilized as the above-mentioned guide rods to accurately insert the fuel rod to the point of insertion. (I.S.)

Control rod drives

International Nuclear Information System (INIS)

Hayakawa, Hiroyasu.

1979-01-01

Purpose: To enable rapid control in a simple circuit by providing a motor control device having an electric capacity capable of simultaneously driving all of the control rods rapidly only in the inserting direction as well as a motor controlling device capable of fine control for the insertion and extraction at usual operation. Constitution: The control rod drives comprise a first motor control device capable of finely controlling the control rods both in inserting and extracting directions, a second motor control device capable of rapidly driving the control rods only in the inserting direction, and a first motor switching circuit and a second motor switching circuit switched by switches. Upon issue of a rapid insertion instruction for the control rods, the second motor switching circuit is closed by the switch and the second motor control circuit and driving motors are connected. Thus, each of the control rod driving motors is driven at a high speed in the inserting direction to rapidly insert all of the control rods. (Yoshino, Y.)

Application of advanced model of radiative heat transfer in a rod geometry to QUENCH and PARAMETER tests

International Nuclear Information System (INIS)

Vasiliev, A.D.; Kobelev, G.V.; Astafieva, V.O.

2007-01-01

Radiative heat transfer is very important in different fields of mechanical engineering and related technologies including nuclear reactors, heat transfer in furnaces, aerospace, different high-temperature assemblies. In particular, in the course of a hypothetical severe accident at PWR-type nuclear reactor the temperatures inside the reactor vessel reach high values at which taking into account of radiative heat exchange between the structures of reactor (including core and other reactor vessel elements) gets important. Radiative heat transfer dominates the late phase of severe accident because radiative heat fluxes (proportional to T4, where T is the temperature) are generally considerably higher than convective and conductive heat fluxes in a system. In particular, heat transfer due to radiation determines the heating and degradation of the core and surrounding steel in-vessel structures and finally influences the composition, temperature and mass of materials pouring out of the reactor vessel after its loss of integrity. Existing models of radiative heat exchange use many limitations and approximations: approximate estimation of view factors and beam lengths; the geometry change in the course of the accident is neglected; the database for emissivities of materials is not complete; absorption/emission by steam-noncondensable medium is taken into account approximately. The module MRAD was developed in this paper to model the radiative heat exchange in rod-like geometry typical of PWR-type reactor. Radiative heat exchange is computed using dividing on zones (zonal method) as in existing radiation models implemented to severe accident numerical codes such as ICARE, SCDAP/RELAP, MELCOR but improved in following aspects: new approach to evaluation of view factors and mean beam length; detailed evaluation of gas absorptivity and emissivity; account of effective radiative thermal conductivity for the large core; account of geometry modification in the course of severe

Fission reactor control rod

International Nuclear Information System (INIS)

Irie, Tomoo.

1991-01-01

The present invention concerns a control rod in a PWR type reactor. A control rod has an inner cladding tube and an outer cladding tube disposed coaxially, and a water draining hole is formed at the inside of the inner cladding tube. Neutron absorbers are filled in an annular gap between the outer cladding tube and the inner cladding tube. The water draining hole opens at the lower end thereof to the top end of the control rod and at the upper end thereof to the side of the upper end plug of the control rod. If the control rod is dropped to a control rod guide thimble for reactor scram, coolants from the control rod guide thimble are flown from the lower end of the water draining hole and discharged from the upper end passing through the water draining hole. In this way, water from the control rod guide thimble is removed easily when the control rod is dropped. Further, the discharging amount of water itself is reduced by the provision of the water draining hole. Accordingly, sufficient control rod dropping speed can be attained. (I.N.)

Fuel rod simulator effects in flooding experiments single rod tests

International Nuclear Information System (INIS)

Nishida, M.

1984-09-01

The influence of a gas filled gap between cladding and pellet on the quenching behavior of a PWR fuel rod during the reflood phase of a LOCA has been investigated. Flooding experiments were conducted with a short length electrically heated single fuel rod simulator surrounded by glass housing. The gap of 0.05 mm width between the Zircaloy cladding and the internal Al 2 O 3 pellets of the rod was filled either wit helium or with argon to vary the radial heat resistance across the gap. This report presents some typical data and an evaluation of the reflood behavior of the fuel rod simulator used. The results show that the quench front propagates faster for increasing heat resistance in the gap between cladding and heat source of the rod. (orig.) [de

Detailed analysis for a control rod worth of the gas turbine high temperature reactor (GTHTR300)

Energy Technology Data Exchange (ETDEWEB)

Nakata, Tetsuo; Katanishi, Shoji; Takada, Shoji; Yan, Xing; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

2002-11-01

GTHTR300 is composed of a simplified and economical power plant based on an inherent safe 600 MWt reactor and a nearly 50% high efficiency gas turbine power conversion cycle. GTHTR300 core consist of annular fuel region, center and outer side reflectors because of cooling it effectively in depressurized accident conditions, and all control rods are located in both side reflectors of annular core. As a thermal neutron spectrum is strongly distorted in reflector regions, an accurate calculation is especially required for the control rod worth evaluation. In this study, we applied the detailed Monte Carlo calculations of a full core model, and confirmed that our design method has enough accuracy. (author)

Influencing factors for early acute cerebrovascular accidents in patients with stroke history following off-pump coronary artery bypass grafting.

Science.gov (United States)

Wang, Bin; Jia, Ming; Jia, Shijie; Wan, Jiuhe; Zhou, Xiao; Luo, Zhimin; Zhou, Ye; Zhang, Jianqun

2014-06-01

To analyse risk factors for early acute cerebrovascular accidents following off-pump coronary artery bypass grafting (OPCAB) in patients with stroke history, and to propose preventive measures to reduce the incidence of these events. A total of 468 patients with a history of stroke underwent OPCAB surgery in Beijing Anzhen Hospital of China from January 2010 to September 2012. They were retrospectively divided into two groups according to the occurrence of early acute cerebrovascular accidents within 48 hours following OPCAB. Multivariate logistic regression analysis was used to find risk or protective factors for early acute cerebrovascular accidents following the OPCAB. Fifty-two patients (11.1%) suffered from early acute cerebrovascular accidents in 468 patients, including 39 cases of cerebral infarction, two cases of cerebral haemorrhage, 11 cases of transient ischaemic attack (TIA). There were significant differences between the two groups in preoperative left ventricular ejection fraction ≤ 35%, severe bilateral carotid artery stenosis, poorly controlled hypertension, intraoperative application of Enclose® II proximal anastomotic device, postoperative acute myocardial infarction, atrial fibrillation, hypotension, ventilation time > 48h, ICU duration >48h and mortality. Multivariate logistic regression analysis showed that preoperative severe bilateral carotid stenosis (OR=6.378, 95%CI: 2.278-20.987) and preoperative left ventricular ejection fraction ≤ 35% (OR=2.737, 95%CI: 1.267-6.389), postoperative acute myocardial infarction (OR=3.644, 95%CI: 1.928-6.876), postoperative atrial fibrillation (OR=3.104, 95%CI:1.135∼8.016) and postoperative hypotension (OR=4.173, 95%CI: 1.836∼9.701) were independent risk factors for early acute cerebrovascular accidents in patients with a history of stroke following OPCAB procedures, while intraoperative application of Enclose® II proximal anastomotic device was protective factor (OR=0.556, 95%CI: 0.337-0.925). This

RODSWELL: a computer code for the thermomechanical analysis of fuel rods under LOCA conditions

International Nuclear Information System (INIS)

Casadei, F.; Laval, H.; Donea, J.; Jones, P.M.; Colombo, A.

1984-01-01

The present report is the user's manual for the computer code RODSWELL developed at the JRC-Ispra for the thermomechanical analysis of LWR fuel rods under simulated loss-of-coolant accident (LOCA) conditions. The code calculates the variation in space and time of all significant fuel rod variables, including fuel, gap and cladding temperature, fuel and cladding deformation, cladding oxidation and rod internal pressure. The essential characteristics of the code are briefly outlined here. The model is particularly designed to perform a full thermal and mechanical analysis in both the azimuthal and radial directions. Thus, azimuthal temperature gradients arising from pellet eccentricity, flux tilt, arbitrary distribution of heat sources in the fuel and the cladding and azimuthal variation of coolant conditions can be treated. The code combines a transient 2-dimensional heat conduction code and a 1-dimentional mechanical model for the cladding deformation. The fuel rod is divided into a number of axial sections and a detailed thermomechanical analysis is performed within each section in radial and azimuthal directions. In the following sections, instructions are given for the definition of the data files and the semi-variable dimensions. Then follows a complete description of the input data. Finally, the restart option is described

Jet behaviors and ejection mode recognition of electrohydrodynamic direct-write

Science.gov (United States)

Zheng, Jianyi; Zhang, Kai; Jiang, Jiaxin; Wang, Xiang; Li, Wenwang; Liu, Yifang; Liu, Juan; Zheng, Gaofeng

2018-01-01

By introducing image recognition and micro-current testing, jet behavior research was conducted, in which the real-time recognition of ejection mode was realized. To study the factors influencing ejection modes and the current variation trends under different modes, an Electrohydrodynamic Direct-Write (EDW) system with functions of current detection and ejection mode recognition was firstly built. Then a program was developed to recognize the jet modes. As the voltage applied to the metal tip increased, four jet ejection modes in EDW occurred: droplet ejection mode, Taylor cone ejection mode, retractive ejection mode and forked ejection mode. In this work, the corresponding relationship between the ejection modes and the effect on fiber deposition as well as current was studied. The real-time identification of ejection mode and detection of electrospinning current was realized. The results in this paper are contributed to enhancing the ejection stability, providing a good technical basis to produce continuous uniform nanofibers controllably.

FRAP-T1: a computer code for the transient analysis of oxide fuel rods

International Nuclear Information System (INIS)

Dearien, J.A.; Miller, R.L.; Hobbins, R.R.; Siefken, L.J.; Baston, V.F.; Coleman, D.R.

1977-02-01

FRAP-T is a FORTRAN IV computer code which can be used to solve for the transient response of a light water reactor (LWR) fuel rod during accident transients such as loss-of-coolant accident (LOCA) or a power-cooling-mismatch (PCM). The coupled effects of mechanical, thermal, internal gas, and material property response on the behavior of the fuel rod are considered. FRAP-T is a modular code with each major computational model isolated within the code and coupled to the main code by subroutine calls and data transfer through argument lists. FRAP-T is coupled to a materials properties subcode (MATPRO) which is used to provide gas, fuel, and cladding properties to the FRAP-T computational subcodes. No material properties need be supplied by the code user. The needed water properties are stored in tables built into the code. Critical heat flux (CHF) and heat transfer correlations for a wide range of coolant conditions are contained in modular subroutines. FRAP-T has been evaluated by making extensive comparisons between predictions of the code and experimental data. Comparison of predicted and experimental results are presented for a range of FRAP-T calculated parameters. The code is presently programmed and running on an IBM-360/75 and a CDC 7600 computer

Maximum/minimum asymmetric rod detection

International Nuclear Information System (INIS)

Huston, J.T.

1990-01-01

This patent describes a system for determining the relative position of each control rod within a control rod group in a nuclear reactor. The control rod group having at least three control rods therein. It comprises: means for producing a signal representative of a position of each control rod within the control rod group in the nuclear reactor; means for establishing a signal representative of the highest position of a control rod in the control rod group in the nuclear reactor; means for establishing a signal representative of the lowest position of a control rod in the control rod group in the nuclear reactor; means for determining a difference between the signal representative of the position of the highest control rod and the signal representative of the position of the lowest control rod; means for establishing a predetermined limit for the difference between the signal representative of the position of the highest control rod and the signal representative of the position of the lowest control rod; and means for comparing the difference between the signals with the predetermined limit. The comparing means producing an output signal when the difference between the signals exceeds the predetermined limit

Effects of thermocouple installation and location on fuel rod temperature measurements

International Nuclear Information System (INIS)

McCormick, R.D.

1983-01-01

This paper describes the results of analyses of nuclear fuel rod cladding temperature data obtained during in-reactor experiments under steady state and transient (simulated loss-of-coolant accident) operating conditions. The objective of the analyses was to determine the effect of thermocouple attachment method and location on measured thermal response. The use of external thermocouples increased the time to critical heat flux (CHF), reduced the blowdown peak temperature, and enhanced rod quench. A comparison of laser welded and resistance welded external thermocouple responses showed that the laser welding technique reduced the indicated cladding steady state temperatures and provided shorter time-to-CHF. A comparison of internal welded and embedded thermocouples indicated that the welded technique gave generally unsatisfactory cladding temperature measurements. The embedded thermocouple gave good, consistent results, but was possibly more fragile than the welded thermocouples. Detailed descriptions of the thermocouple designs, attachment methods and locations, and test conditions are provided

Interaction of radionuclides in severe accident conditions

International Nuclear Information System (INIS)

Nagrale, Dhanesh B.; Bera, Subrata; Deo, Anuj Kumar; Paul, U.K.; Prasad, M.; Gaikwad, A.J.

2015-01-01

Nuclear power plants are designed with inherent engineering safety systems and associated operational procedures that provide an in-depth defence against accidents. Radionuclides such as Iodine, Cesium, Tellurium, Barium, Strontium, Rubidium, Molybdenum and many others may get released during a severe accident. Among these, Iodine, one of the fission products, behaviour is significant for the analysis of severe accident consequences because iodine is a chemically more active to the potential components released to the environment. During severe accident, Iodine is released and transported in aqueous, organic and inorganic forms. Iodine release from fuel, iodine transport in primary coolant system, containment, and reaction with control rods are some of the important phases in a severe accident scenario. The behaviour of iodine is governed by aerosol physics, depletion mechanisms gravitational settling, diffusiophoresis and thermophoresis. The presence of gaseous organic compounds and oxidizing compounds on iodine, reactions of aerosol iodine with boron and formation of cesium iodide which results in more volatile iodine release in containment play significant roles. Water radiolysis products due to presence of dissolved impurities, chloride ions, organic impurities should be considered while calculating iodine release. Containment filtered venting system (CFVS) consists of venturi scrubber and a scrubber tank which is dosed with NaOH and NaS_2O_3 in water where iodine will react with the chemicals and convert into NaI and Na_2SO_4. This paper elaborates the issues with respect to interaction of radionuclides and its consideration in modeling of severe accident. (author)

Evaluation of the behavior of waterlogged fuel rod failures in LWRs

International Nuclear Information System (INIS)

Siegel, B.

1977-11-01

A summary of the available information on waterlogged fuel rod failures is presented. The information includes experimental results from waterlogging tests in research reactors, observations of waterlogging failures in commercial reactors, and reactor vendor assessments. It is concluded that (a) operating restrictions to reduce pellet/cladding interactions also reduce the potential for waterlogging failures during transients, (b) tests to simulate accident conditions produced the worst waterlogging failures, and (c) there is no apparent threat from waterlogging failures to the overall coolability of the core or to safe reactor shutdown

Radiation accident of 60Co contamination. Mexico 1984

International Nuclear Information System (INIS)

1985-01-01

The action taken to mitigate the consequences of a radiation accident occuring in the State of Chihuahua, Mexico, is described, when a no longer in use cobalt-60 teletherapy unit, with radioactive pellets still inside, was sold, unwittingly as ordinary scrap to be finally made into reinforced steel rods. The finished metallic items, unknowingly contaminated with cobalt-60, were subsequently sold in central and northern Mexico and in the United States of America. The junkyard, transport vehicle, several foundries and some streets of two cities and the road between them were also made radioactive by the accident. The discovery of and search for the radioactive metallic products is described as is their final disposal and the decontamination of the affected sites. Individual and collective radiation doses is estimated. (author)

Control rod drives

International Nuclear Information System (INIS)

Hayakawa, Hiroyasu; Kawamura, Atsuo.

1979-01-01

Purpose: To reduce pellet-clad mechanical interactions, as well as improve the fuel safety. Constitution: In the rod drive of a bwr type reactor, an electric motor operated upon intermittent input such as of pulse signals is connected to a control rod. A resolver for converting the rotational angle of the motor to electric signals is connected to the rotational shaft of the motor and the phase difference between the output signal from the resolver and a reference signal is adapted to detect by a comparator. Based on the detection result, the controller is actuated to control a motor for control rod drive so that fine control for the movement of the control rod is made possible. This can reduce the moving distance of the control rod, decrease the thermal stress applied to the control rod and decrease the pellet clad mechanical interaction failures due to thermal expansion between the cladding tube and the pellets caused by abrupt changes in the generated power. (Furukawa, Y.)

Ejection experience in Serbian air force, 1990-2010

Directory of Open Access Journals (Sweden)

Pavlović Miroslav

2014-01-01

Full Text Available Background/Aim. Ejection injuries are the problem for air forces. The present risk for injuries is still too high, approximately 30-50%. This study was an effort to determine factors responsible for and contributing to injuries in the Serbian Air Force (SAF in the last two decades. Methods. All ejection cases in the SAF between 1990 and 2010 were analyzed. The collected data were: aircraft type, ejection seat generation, pilots ´ age and experience, causes of ejection, aeronautical parameters, the condition of aircraft control and types of injuries. For ease of comparison the U.S. Air Force Safety Regulation was used to define of major injuries: hospitalization for 5 days or more, loss of consciousness for over 5 min, bone fracture, joint dislocation, injury to any internal organ, any third-degree burn, or second-degree burn over 5% of the body surface area. Results. There were 52 ejections (51 pilots and 1 mechanic on 44 airplanes. The ejected persons were from 22 to 46 years, average 32 years. Major injuries were present in 25.49% cases. Of all the ejected pilots 9.61% had fractures of thoracic spine, 11.53% fractures of legs, 3.48% fractures of arms. Of all major injuries, fractures of thoracic spine were 38.46%. None of the pilots had experienced ejection previously. Conclusion. Our results suggest to obligatory take preventive measures: magnetic resonance imaging (MRI scan must be included in the standard pilot selection procedure and procedure after ejection. Physical conditioning of pilots has to be improved. Training on ejection trainer has to be accomplished, too.

Subchannel Scale Thermal-Hydraulic Analysis of Rod Bundle Geometry under Single-phase Adiabatic Conditions Using CUPID

Energy Technology Data Exchange (ETDEWEB)

Yoon, Seok Jong; Park, Goon Cherl; Cho, Hyoung Kyu [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In Korea, subchannel analysis code, MATRA has been developed by KAERI (Korea Atomic Energy Research Institute). MATRA has been used for reactor core T/H design and DNBR (Departure from Nucleate Boiling Ratio) calculation. Also, the code has been successfully coupled with neutronics code and fuel analysis code. However, since major concern of the code is not the accident simulation, some features of the code are not optimized for the accident conditions, such as the homogeneous model for two-phase flow and spatial marching method for numerical scheme. For this reason, in the present study, application of CUPID for the subchannel scale T/H analysis in rod bundle geometry was conducted. CUPID is a component scale T/H analysis code which adopts three dimensional two-fluid three-field model developed by KAERI. In this paper, the validation results of the CUPID code for subchannel scale rod bundle analysis at single phase adiabatic conditions were presented. At first, the physical models required for a subchannel scale analysis were implemented to CUPID. In the future, the scope of validation tests will be extended to diabetic and two phase flow conditions and required models will be implemented into CUPID.

Status of rod consolidation

International Nuclear Information System (INIS)

Bailey, W.J.

1985-04-01

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 10 0 C) 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

Considerations relating to the presence of water in the reactor cavity during severe accidents

International Nuclear Information System (INIS)

Perez, F.; Morales, M.D.

1994-01-01

The purpose of this paper is to present some of the factors, both positive and negative, associated with the presence of water in the reactor cavity. The presence of water in the reactor cavity is one of the factors whose influence on the evolution of severe accidents must be determined since, on the one hand, it has an impact on some of the most significant severe accident phenomena and, on the other, it could be an important factor when preparing accident management strategies resulting from containment analyses. In spite of the initial intuitive impression that water in the reactor cavity must always be beneficial, certain phenomena, such as the following must also be taken into account before developing accident management strategies: - Higher production of steam - Possibility of steam explosions - Increased production of H 2 due to oxidation of steel components of the melted core ejected from the vessel - More oxidation energy released due to the presence of oxygen in the cavity (Author)

Safety analysis results for the control rod banks withdrawal event at a full power of the SMART-P

International Nuclear Information System (INIS)

Yang, S. H.; Chung, Y. J.; Kim, H. C.; Zee, S. Q.

2005-01-01

For the validation of the 330 MWt SMART (System-integrated Modular Advanced ReacTor), a detailed design for the SMART-P has been accomplished by KAERI. In the SMART-P design similar to the SMART design, the soluble boron free design is adapted. This concept results in a larger reactivity worth of the control rod bank compared to that of the commercial pressurized water reactor. Moreover, in the SMART-P design, the control rod banks are fairly well inserted into the core, even at a full power condition. Therefore, accidents related to the reactivity anomalies have been evaluated as crucial events when compared to the other initiating events. In this paper, safety analysis for the control rod banks withdrawal event at a full power of the SMART-P has been accomplished by considering various initial conditions, different withdrawal times of the control rod banks and the reactivity feedback. To perform the safety analysis, the TASS/SMR (Transients And Setpoint Simulation/Small and Medium Reactor) code for a system response and SSF-1 correlation for a CHFR (Critical Heat Flux Ratio) have been used

Ultrasound - Aided ejection in micro injection molding

Science.gov (United States)

Masato, D.; Sorgato, M.; Lucchetta, G.

2018-05-01

In this work, an ultrasound-aided ejection system was designed and tested for different polymers (PS, COC and POM) and mold topographies. The proposed solution aims at reducing the ejection friction by decreasing the adhesion component of the frictional force, which is controlled by the contact area developed during the filling stage of the injection molding process. The experimental results indicate a positive effect of ultrasound vibration on the friction force values, with a maximum reduction of 16. Moreover, it is demonstrated that the ultrasound effect is strictly related to both polymer selection and mold roughness. The combined effect on the ejection force of mold surface roughness, melt viscosity during filling and polymer elastic modulus at ejection was modeled to the experimental data, in order to demonstrate that the effect of ultrasound vibration on the ejection friction reduction is due to the heating of the contact interface and the consequent reduction of the polymer elastic modulus.

Thermalydraulic processes in the reactor coolant system of a BWR under severe accident conditions

International Nuclear Information System (INIS)

Hodge, S.A.

1990-01-01

Boiling water reactors (BWRs) incorporate many unique structural features that make their expected response under severe accident conditions very different from that predicted in the case of pressurized water reactor accident sequences. Automatic main steam isolation valve (MIV) closure as the vessel water level approaches the top of the core would cause reactor vessel isolation while automatic recirculation pump trip would limit the in-vessel flows to those characteristic of natural circulation (as disturbed by vessel relief valve actuation). This paper provides a discussion of the BWR control blade, channel box, core plate, control rod guide tube, and reactor vessel safety relief valve (SRV) configuration and the effects of these structural components upon thermal hydraulic processes within the reactor vessel under severe accident conditions. The dominant BWR severe accident sequences as determined by probabilistic risk assessment are described and the expected timing of events for the unmitigated short-term station blackout severe accident sequence at the Peach Bottom atomic power station is presented

Control rod drive

International Nuclear Information System (INIS)

Okutani, Tetsuro.

1988-01-01

Purpose: To provide a simple and economical control rod drive using a control circuit requiring no pulse circuit. Constitution: Control rods in a BWR type reactor are driven by hydraulic pressure and inserted or withdrawn in the direction of applying the hydraulic pressure. The direction of the hydraulic pressure is controlled by a direction control valve. Since the driving for the control rod is extremely important in view of the operation, a self diagnosis function is disposed for rapid inspection of possible abnormality. In the present invention, two driving contacts are disposed each by one between the both ends of a solenoid valve of the direction control valve for driving the control rod and the driving power source, and diagnosis is conducted by alternately operating them. Therefore, since it is only necessary that the control circuit issues a driving instruction only to one of the two driving contacts, the pulse circuit is no more required. Further, since the control rod driving is conducted upon alignment of the two driving instructions, the reliability of the control rod drive can be improved. (Horiuchi, T.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2004-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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.

Control rod withdrawal monitoring device

International Nuclear Information System (INIS)

Ebisuya, Mitsuo.

1984-01-01

Purpose: To prevent the power ramp even if a plurality of control rods are subjected to withdrawal operation at a time, by reducing the reactivity applied to the reactor. Constitution: The control rod withdrawal monitoring device is adapted to monitor and control the withdrawal of the control rods depending on the reactor power and the monitoring region thereof is divided into a control rod group monitoring region a transition region and a control group monitoring not interfere region. In a case if the distance between a plurality of control rods for which the withdrawal positions are selected is less than a limiting value, the coordinate for the control rods, distance between the control rods and that the control rod distance is shorter are displayed on a display panel, and the withdrawal for the control rods are blocked. Accordingly, even if a plurality of control rods are subjected successively to the withdrawal operation contrary to the control rod withdrawal sequence upon high power operation of the reactor, the power ramp can be prevented. (Kawakami, Y.)

Space weather and coronal mass ejections

CERN Document Server

Howard, Tim

2013-01-01

Space weather has attracted a lot of attention in recent times. Severe space weather can disrupt spacecraft, and on Earth can be the cause of power outages and power station failure. It also presents a radiation hazard for airline passengers and astronauts. These ""magnetic storms"" are most commonly caused by coronal mass ejections, or CMES, which are large eruptions of plasma and magnetic field from the Sun that can reach speeds of several thousand km/s. In this SpringerBrief, Space Weather and Coronal Mass Ejections, author Timothy Howard briefly introduces the coronal mass ejection, its sc

Enhanced accident-tolerant fuel (EATF)

International Nuclear Information System (INIS)

Strumpell, John

2013-01-01

The Fukushima accident provided a strong reminder that the exothermic reaction between zirconium and steam, and the attendant hydrogen generation, can significantly affect the course of a severe accident. Part of the response to the accident was increased interest in the extent to which the fuel itself can mitigate the consequences of a severe accident. Improved fuel alone is not sufficient to provide the desired increase in reactor safety, but it can provide an important contribution. With support from the US Department of Energy, AREVA has brought together a team that includes researchers (AREVA, Electric Power Research Institute, Savannah River National Laboratory, University of Florida, and University of Wisconsin), a fuel vendor (AREVA), and utilities (Duke Energy and Tennessee Valley Authority). The goal of the project is to develop new technologies that can be deployed in a lead assembly within ten years. The researchers have proposed a variety of approaches for improving the performance of the fuel, including new cladding and structural materials, fuel pellets with improved thermal characteristics, and coatings on the fuel rods. The expected performance of fuels that apply these technologies will be judged against the requirements of the vendor and utilities to determine those that are most promising for immediate development and those that may be suited for development in the future. The first review will consider the manufacturability of the proposed designs; the second will focus on performance. Materials that are suitable for immediate development will be considered for irradiation in a test reactor and subsequent use in lead assembly designs

Development of Severe Accident Containment Analysis Package

Energy Technology Data Exchange (ETDEWEB)

Park, Chang-Hwan; Kim, Dong-Min; Seo, Jea-Uk; Lee, Dea-Young; Park, Soon-Ho; Lee, Jae-Gwon; Lee, Jin-Yong; Lee, Byung-Chul [FNC Technology Co., Yongin (Korea, Republic of)

2016-10-15

In safety viewpoint, the pressure and temperature of the containment is the important parameters, of course, the local hydrogen concentration is also the parameter of the major concern because of its flammability and the risk of the detonation. In addition, there are possibilities of occurrence of other relevant phenomena following the reactor core melting such as DCH(direct containment heating) due to HPME(high pressure melt ejection), steam explosion due to fuel-coolant interaction in the reactor cavity and molten core concrete interaction at the late stage. It is important to predict the containment responses during a severe accident by a reasonable accuracy for establishing of effective mitigation strategies and preparation of the safety features required. In this paper, the overview of the SACAP development status is presented. SACAP is developed so as to be able to analyze, so called, Ex-Vessel severe accident phenomena including thermal-hydraulics, combustible gas burn, direct containment heating, steam explosion and molten core-concrete interaction. At the parallel time, SACAP and In-Vessel analysis module named CSPACE are processed for integration through MPI communication coupling. Development of the integrated severe accident analysis code system will be completed in following one year to make the code revision zero to be released.

Mass ejections from the solar corona into interplanetary space

International Nuclear Information System (INIS)

Hildner, E.

1977-01-01

Mass ejections from the corona are common occurrances, as observations with the High Altitude Observatory's white light coronagraph aboard Skylab showed. During 227 days of operation in 1973 and 1974 at least 77 mass ejections were observed and as many more probably occurred unobserved. It is suggested that the frequency of ejections varies with the solar cycle and that ejections may contribute 10 percent or more of the total solar mass efflux to the interplanetary medium at solar maximum. Since ejections are confined to relatively low latitudes, their fractional mass flux contribution is greater near the ecliptic than far from it. From the behavior of ejecta, we can estimate the magnitude of the force driving them through the corona. It is also suggested that loop-shaped ejection - the largest fraction of ejections - are driven, primarily, by magnetic forces. By comparison, gas pressure forces are negligible, and forces due to wave pressure are completely inadequate. That magnetic forces are important is consistent with observation that ejections seem to come, primarily, from regions where the magnetic field is more intense and more complex than elsewhere. Indeed, ejections are associated with phenomena (flares and eruptive prominences) which occur over lines separating regions of opposite polarities. (Auth.)

Results of the first nuclear blowdown test on single fuel rods (LOC-11 Series in PBF)

International Nuclear Information System (INIS)

Larson, J.R.; Evans, D.R.; McCardell, R.K.

1978-01-01

This paper presents results of the first nuclear blowdown tests (LOC-11A, LOC-11B, LOC-11C) ever conducted. The Loss-of-Coolant Accident (LOCA) Test Series is being conducted in the Power Burst Facility (PBF) reactor at the Idaho National Engineering Laboratory, near Idaho Falls, Idaho, for the Nuclear Regulatory Commission. The objective of the LOC-11 tests was to obtain data on the behavior of pressurized and unpressurized rods when exposed to a blowdown similar to that expected in a pressurized water reactor (PWR) during a hypothesized double-ended cold-leg break. The data are being used for the development and verification of analytical models that are used to predict coolant and fuel rod pressure during a LOCA in a PWR

Thermal behavior simulation of a nuclear fuel rod through an eletrically heated rod

International Nuclear Information System (INIS)

Lima, R. de C.F. de.

1984-01-01

In thermalhydraulic loops the nuclear industry often uses electrically heated rods to simulate power transients, which occur in nuclear fuel rods. The development and design of a electrically heated rod, by supplying the dimensions and materials which should be used in order to yeld the same temperature and heat flux at the surfaces of the nuclear rod and the electrically heated rod are presented. To a given nuclear transient this equality was obtained by fitting the linear power through the lumped parameters technique. (Author) [pt

SSYST, a code-system for analysing transient LWR fuel rod behaviour under off-normal conditions

International Nuclear Information System (INIS)

Borgwaldt, H.; Gulden, W.

1983-01-01

SSYST is a code-system for analysing transient fuel rod behaviour under off-normal conditions, developed conjointly by the Institut fuer Kernenergetik und Energiesysteme (IKE), Stuttgart, and Kernforschungszentrum Karlsruhe (KfK) under contract of Projek Nukleare Sicherheit (PNS) at KfK. The main differences between SSYST and similar codes are (1) an open-ended modular code organisation, and (2) a preference for simple models, wherever possible. While the first feature makes SSYST a very flexible tool, easily adapted to changing requirements, the second feature leads to short execution times. The analysis of transient rod behaviour under LOCA boundary conditions takes 2 min cpu-time (IBM-3033), so that extensive parametric studies become possible. This paper gives an outline of the overall code organisation and a general overview of the physical models implemented. Besides explaining the routine application of SSYST in the analysis of loss-of-coolant accidents, examples are given of special applications which have led to a satisfactory understanding of the decisive influence of deviations from rotational symmetry on the fuel rod perimeter. (author)

SSYST: A code-system for analyzing transient LWR fuel rod behaviour under off-normal conditions

International Nuclear Information System (INIS)

Borgwaldt, H.; Gulden, W.

1983-01-01

SSYST is a code-system for analyzing transient fuel rod behaviour under off-normal conditions, developed conjointly by the Institut fur Kernenergetik und Energiesysteme (IKE), Stuttgart, and Kernforschungszentrum Karlsruhe (KfK) under contract of Projekt Nukleare Sicherheit (PNS) at KfK. The main differences between SSYST and similar codes are an open-ended modular code organization, and a preference for simple models, wherever possible. While the first feature makes SSYST a very flexible tool, easily adapted to changing requirements, the second feature leads to short execution times. The analysis of transient rod behaviour under LOCA boundary conditions takes 2 min cpu-time (IBM-3033), so that extensive parametric studies become possible. This paper gives an outline of the overall code organisation and a general overview of the physical models implemented. Besides explaining the routine application of SSYST in the analysis of loss-of-coolant accidents, examples are given of special applications which have led to a satisfactory understanding of the decisive influence of deviations from rotational symmetry on the fuel rod perimeter

Supernova mass ejection and core hydrodynamics

International Nuclear Information System (INIS)

Colgate, S.A.

1978-01-01

Simplifications that have emerged in the descriptions of stellar unstable collapse to a neutron star are discussed. The neutral current weak interaction leads to almost complete neutrino trapping in the collapse and to an electron fraction Y/sub e/ congruent to 0.35 in equilibrium with trapped electron neutrinos and ''iron'' nuclei. A soft equation of state (γ congruent to 1.30) leads to collapse, and bounce occurs on a hard core, γ = 2.5, at nuclear densities. Neutrino emission is predicted from a photosphere at r congruent to 2 x 10 7 cm and E/sub ν/ congruent to 10 MeV. The ejection of matter by an elastic core bounce and a subsequent escaping shock is marginal and may not be predicted for accurate values of the equation of state. A new concept of Rayleigh-Taylor driven core instabilities is invoked to predict an increased mass ejection either due to an increased flux and energy of neutrinos at second bounce time and, or, the rapid 0.1 to 0.4 second formation of a more energetically bound neutron star. The instability is caused by highly neutronized external matter from which neutrinos have escaped being supported by lighter matter of the lepton trapped core. An initial anisotropy of 10 -2 to 10 -3 should lead to adequately rapid (several milliseconds) overturn following several (2 to 4) bounces. Subsequent to the overturnwith or without a strong ejection shock, a weak ejection shock will allow an accretion shock to form on the ''cold'' neutron star core due to the reimplosion or rarefaction wave in the weakly ejected matter. The accretion shock forms at low enough mass accumulation rate, 1 / 2 M/sub solar/ sec -1 , such that a black body neutrino flux can escape from the shock front (kT congruent to 10 MeV, [E/sub ν/] congruent to 30 MeV). This strongly augments the weaker bounce ejection shock by heating the external matter in the mantle by electron neutrino scattering (congruent to 10 52 ergs) causing adequate mass ejection

Nuclear fuel rod loading apparatus

International Nuclear Information System (INIS)

King, H.B.; Macivergan, R.; Mckenzie, G.W.

1980-01-01

An apparatus incorporating a microprocessor control is provided for automatically loading nuclear fuel pellets into fuel rods commonly used in nuclear reactor cores. The apparatus comprises a split ''v'' trough for assembling segments of fuel pellets in rows and a shuttle to receive the fuel pellets from the split ''v'' trough when the two sides of the split ''v'' trough are opened. The pellets are weighed while in the shuttle, and the shuttle then moves the pellets into alignment with a fuel rod. A guide bushing is provided to assist the transfer of the pellets into the fuel rod. A rod carousel which holds a plurality of fuel rods presents the proper rod to the guide bushing at the appropriate stage in the loading sequence. The bushing advances to engage the fuel rod, and the shuttle advances to engage the guide bushing. The pellets are then loaded into the fuel rod by a motor operated push rod. The guide bushing includes a photocell utilized in conjunction with the push rod to measure the length of the row of fuel pellets inserted in the fuel rod

Control rod shutdown system

International Nuclear Information System (INIS)

Miyamoto, Yoshiyuki; Higashigawa, Yuichi.

1996-01-01

The present invention provides a control rod terminating system in a BWR type nuclear power plant, which stops an induction electric motor as rapidly as possible to terminate the control rods. Namely, the control rod stopping system controls reactor power by inserting/withdrawing control rods into a reactor by driving them by the induction electric motor. The system is provided with a control device for controlling the control rods and a control device for controlling the braking device. The control device outputs a braking operation signal for actuating the braking device during operation of the control rods to stop the operation of the control rods. Further, the braking device has at least two kinds of breaks, namely, a first and a second brakes. The two kinds of brakes are actuated by receiving the brake operation signals at different timings. The brake device is used also for keeping the control rods after the stopping. Even if a stopping torque of each of the breaks is small, different two kinds of brakes are operated at different timings thereby capable of obtaining a large stopping torque as a total. (I.S.)

Fabrication and Application of Mono-sized Spherical Micro Particles by Pulsated Orifice Ejection Method

Directory of Open Access Journals (Sweden)

DONG Wei

2018-02-01

Full Text Available A novel technology called pulsated orifice ejection method(POEM and used for preparing mono-sized and high-precision spherical micro particles was introduced in this article. The working principle of the technique was illustrated and it was in two modes:low-melting point diaphragm mode and high-melting point rod mode, depending on the different melting points of materials. The particles prepared by POEM have the advantages of mono-sized, uniform and controllable particle size, high sphericity, and consistent thermal history. By introducing the application of particles prepared by this method, showing the huge application prospects of this technology in electronic packaging, bioengineering, micro-fabrication, rapid solidification analysis of metal droplets, additive manufacturing and so on.With the development of POEM, this technology is predicted to have wider prospects due to its unique characteristics.

Austrian contributions to fuel rod failure models shown at the International Standard Problem ISP-14

International Nuclear Information System (INIS)

Sdouz, G.

1984-04-01

The computer code BALON-2A was improved to perform the International Standard Problem ISP-14. The main extensions are the implementation of input-options and the development of a model to predict the pressure in the fuel rod gap. With these improvements and some calculations for input values satisfying results have been obtained. This is remarkable because loss of coolant accident analyses are performed usually with larger computer codes. (Author) [de

Fuel rod leak detector

International Nuclear Information System (INIS)

Womack, R.E.

1978-01-01

A typical embodiment of the invention detects leaking fuel rods by means of a radiation detector that measures the concentration of xenon-133 ( 133 Xe) within each individual rod. A collimated detector that provides signals related to the energy of incident radiation is aligned with one of the ends of a fuel rod. A statistically significant sample of the gamma radiation (γ-rays) that characterize 133 Xe is accumulated through the detector. The data so accumulated indicates the presence of a concentration of 133 Xe appropriate to a sound fuel rod, or a significantly different concentration that reflects a leaking fuel rod

Thermal phenomenae in nuclear fuel rods

International Nuclear Information System (INIS)

Baigorria, Carlos.

1983-12-01

Thermal phenomenae occurring in a nuclear fuel rod under irradiation are studied. The most important parameters of either steady or transient thermal states are determined. The validity of applying the Fourier's approximation equations to these problems is also studied. A computer program TRANS is developed in order to study the transient cases. This program solves a system of coupled, non-linear partial differential equations, of parabolic type, in cylindrical coordinates with various boundary conditions. The benchmarking of the TRANS program is done by comparing its predictions with the analytical solution of some simplified transient cases. Complex transient cases such as those corresponding to characteristic reactor accidents are studied, in particular for typical pressurized heavy water reactor (PHWR) fuel rods, such as those of Atucha I. The Stefan problem emerging in the case of melting of the fuel element is solved. Qualitative differences between the classical Stefan problem, without inner sources, and that one, which includes sources are discussed. The MSA program, for solving the Stefan problem with inner sources is presented; and furthermore, it serves to predict thermal evolution, when the fuel element melts. Finally a model for fuel phase change under irradiation is developed. The model is based on the dimensional invariants of the percolation theory when applied to the connectivity of liquid spires nucleated around each fission fragment track. Suggestions for future research into the subject are also presented. (autor) [es

Safety against releases in severe accidents. Final report

International Nuclear Information System (INIS)

Lindholm, I.; Berg, Oe.; Nonboel, E.

1997-12-01

The work scope of the RAK-2 project has involved research on quantification of the effects of selected severe accident phenomena for Nordic nuclear power plants, development and testing of a computerised accident management support system and data collection and description of various mobile reactors and of different reactor types existing in the UK. The investigations of severe accident phenomena focused mainly on in-vessel melt progression, covering a numerical assessment of coolability of a degraded BWR core, the possibility and consequences of a BWR reactor to become critical during reflooding and the core melt behavior in the reactor vessel lower plenum. Simulant experiments were carried out to investigate lower head hole ablation induced by debris discharge. In addition to the in-vessel phenomena, a limited study on containment response to high pressure melt ejection in a BWR and a comparative study on fission product source term behaviour in a Swedish PWR were performed. An existing computerised accident management support system (CAMS) was further developed in the area of tracking and predictive simulation, signal validation, state identification and user interface. The first version of a probabilistic safety analysis module was developed and implemented in the system. CAMS was tested in practice with Barsebaeck data in a safety exercise with the Swedish nuclear authority. The descriptions of the key features of British reactor types, AGR, Magnox, FBR and PWR were published as data reports. Separate reports were issued also on accidents in nuclear ships and on description of key features of satellite reactors. The collected data were implemented in a common Nordic database. (au)

On the mass ejected by supernova explosions

International Nuclear Information System (INIS)

Bohigas, J.

1984-01-01

A simple model is developed in order to calculate the mass ejected by superonovae. We find that the 185, 1006, 1572 and 1604 AD events, all of them classified as either probable or possible type I supernovae, ejected between 0.1 and 0.4 solar masses with an expansion velocity of roughly 10,000 km s -1 . This range of masses suggests that a collapsed object is at the center of the remnants produced by these supernovae if the precursor was a white dwarf whose mass was closed to the Chandrasekhar limit. For the Crab we obtain an ejected mass of 0.45 Msub(sun) and point out that this value is not in contradiction with a proposal in which the moderate helium stars are good candidates for producing this kind of supernovae. Finally we obtain an ejected mass of 3.1 Msub(sun) for Cas A, indicating that a type II event produced this remnant. This ejected mass is closed to what would be expected for a progenitor like an OBN star. (author)

Fuel rods

International Nuclear Information System (INIS)

Hattori, Shinji; Kajiwara, Koichi.

1980-01-01

Purpose: To ensure the safety for the fuel rod failures by adapting plenum springs to function when small forces such as during transportation of fuel rods is exerted and not to function the resilient force when a relatively great force is exerted. Constitution: Between an upper end plug and a plenum spring in a fuel rod, is disposed an insertion member to the lower portion of which is mounted a pin. This pin is kept upright and causes the plenum spring to function resiliently to the pellets against the loads due to accelerations and mechanical vibrations exerted during transportation of the fuel rods. While on the other hand, if a compression force of a relatively high level is exerted to the plenum spring during reactor operation, the pin of the insertion member is buckled and the insertion member is inserted to the inside of the plenum spring, whereby the pellets are allowed to expand freely and the failures in the fuel elements can be prevented. (Moriyama, K.)

Assessment Of Source Term And Radiological Consequences For Design Basis Accident And Beyond Design Basis Accident Of The Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong; Tran Tri Vien

2011-01-01

The paper presents results of the assessment of source terms and radiological consequences for the Design Basis Accident (DBA) and Beyond Design Basis Accident (BDBA) of the Dalat Nuclear Research Reactor. The dropping of one fuel assembly during fuel handling operation leading to the failure of fuel cladding and the release of fission products into the environment was selected as a DBA for the analysis. For the BDBA, the introduction of a step positive reactivity due to the falling of a heavy block from the rotating bridge crane in the reactor hall onto a part of the platform where are disposed the control rod drives is postulated. The result of the radiological consequence analyses shows that doses to members of the public are below annual dose limit for both DBA and BDBA events. However, doses from exposure to operating staff and experimenters working inside the reactor hall are predicted to be very high in case of BDBA and therefore the protective actions should be taken when the accident occurs. (author)

Severe Accident Recriticality Analyses (SARA)

Energy Technology Data Exchange (ETDEWEB)

Frid, W. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Hoejerup, F. [Risoe National Lab. (Denmark); Lindholm, I.; Miettinen, J.; Puska, E.K. [VTT Energy, Helsinki (Finland); Nilsson, Lars [Studsvik Eco and Safety AB, Nykoeping (Sweden); Sjoevall, H. [Teoliisuuden Voima Oy (Finland)

1999-11-01

Recriticality in a BWR has been studied for a total loss of electric power accident scenario. In a BWR, the B{sub 4}C control rods would melt and relocate from the core before the fuel during core uncovery and heat-up. If electric power returns during this time-window unborated water from ECCS systems will start to reflood the partly control rod free core. Recriticality might take place for which the only mitigating mechanisms are the Doppler effect and void formation. In order to assess the impact of recriticality on reactor safety, including accident management measures, the following issues have been investigated in the SARA project: 1. the energy deposition in the fuel during super-prompt power burst, 2. the quasi steady-state reactor power following the initial power burst and 3. containment response to elevated quasi steady-state reactor power. The approach was to use three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto 1 plant in Finland with all three codes. The core state initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality - both superprompt power bursts and quasi steady-state power generation - for the studied range of parameters, i. e. with core uncovery and heat-up to maximum core temperatures around 1800 K and water flow rates of 45 kg/s to 2000 kg/s injected into the downcomer. Since the recriticality takes place in a small fraction of the core the power densities are high which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal/g, was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding

Severe Accident Recriticality Analyses (SARA)

International Nuclear Information System (INIS)

Frid, W.; Hoejerup, F.; Lindholm, I.; Miettinen, J.; Puska, E.K.; Nilsson, Lars; Sjoevall, H.

1999-11-01

Recriticality in a BWR has been studied for a total loss of electric power accident scenario. In a BWR, the B 4 C control rods would melt and relocate from the core before the fuel during core uncovery and heat-up. If electric power returns during this time-window unborated water from ECCS systems will start to reflood the partly control rod free core. Recriticality might take place for which the only mitigating mechanisms are the Doppler effect and void formation. In order to assess the impact of recriticality on reactor safety, including accident management measures, the following issues have been investigated in the SARA project: 1. the energy deposition in the fuel during super-prompt power burst, 2. the quasi steady-state reactor power following the initial power burst and 3. containment response to elevated quasi steady-state reactor power. The approach was to use three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto 1 plant in Finland with all three codes. The core state initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality - both superprompt power bursts and quasi steady-state power generation - for the studied range of parameters, i. e. with core uncovery and heat-up to maximum core temperatures around 1800 K and water flow rates of 45 kg/s to 2000 kg/s injected into the downcomer. Since the recriticality takes place in a small fraction of the core the power densities are high which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal/g, was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding

Multiple fuel rod gripper

International Nuclear Information System (INIS)

Shields, E.P.

1987-01-01

An apparatus is described for gripping an array of rods comprising: (a) gripping members grippingly engageable with the rods, each of which has a hollow portion terminating in an open end for receiving the end of one of the rods; (b) a closing means for causing the hollow portion of each of the gripping members to apply substantially the same gripping force onto the end of its respective rod, including (i) a locking plate having a plurality of tapered holes registrable with the array of rods, wherein the exterior of each of the gripping members is tapered and nested within one of the tapered holes, (ii) a withdrawing means having a hydraulic plunger operatively connected to each of the gripping members for applying a substantially identical withdrawing force on each of the gripping members, whereby the hollow portion of each of the gripping members applies substantially the same gripping force on its respective rod, and (c) means for detecting whether each of the gripping members has grippingly engaged its respective rod

High-yield production of hydrophobins RodA and RodB from Aspergillus fumigatus in Pichia pastoris

DEFF Research Database (Denmark)

Pedersen, Mona Højgaard; Borodina, Irina; Moresco, Jacob Lange

2011-01-01

A as well as rRodB were able to convert a glass surface from hydrophilic to hydrophobic similar to native RodA, but only rRodB was able to decrease the hydrophobicity of a Teflon-like surface to the same extent as native RodA, while rRodA showed this ability to a lesser extent. Recombinant RodA and native...

Coronal Mass Ejections An Introduction

CERN Document Server

Howard, Timothy

2011-01-01

In times of growing technological sophistication and of our dependence on electronic technology, we are all affected by space weather. In its most extreme form, space weather can disrupt communications, damage and destroy spacecraft and power stations, and increase radiation exposure to astronauts and airline passengers. Major space weather events, called geomagnetic storms, are large disruptions in the Earth’s magnetic field brought about by the arrival of enormous magnetized plasma clouds from the Sun. Coronal mass ejections (CMEs) contain billions of tons of plasma and hurtle through space at speeds of several million miles per hour. Understanding coronal mass ejections and their impact on the Earth is of great interest to both the scientific and technological communities. This book provides an introduction to coronal mass ejections, including a history of their observation and scientific revelations, instruments and theory behind their detection and measurement, and the status quo of theories describing...

Cone dystrophy with "supernormal" rod ERG: psychophysical testing shows comparable rod and cone temporal sensitivity losses with no gain in rod function.

Science.gov (United States)

Stockman, Andrew; Henning, G Bruce; Michaelides, Michel; Moore, Anthony T; Webster, Andrew R; Cammack, Jocelyn; Ripamonti, Caterina

2014-02-10

We report a psychophysical investigation of 5 observers with the retinal disorder "cone dystrophy with supernormal rod ERG," caused by mutations in the gene KCNV2 that encodes a voltage-gated potassium channel found in rod and cone photoreceptors. We compared losses for rod- and for cone-mediated vision to further investigate the disorder and to assess whether the supernormal ERG is associated with any visual benefit. L-cone, S-cone, and rod temporal acuity (critical flicker fusion frequency) were measured as a function of target irradiance; L-cone temporal contrast sensitivity was measured as a function of temporal frequency. Temporal acuity measures revealed that losses for vision mediated by rods, S-cones, and L-cones are roughly equivalent. Further, the gain in rod function implied by the supernormal ERG provides no apparent benefit to near-threshold rod-mediated visual performance. The L-cone temporal contrast sensitivity function in affected observers was similar in shape to the mean normal function but only after the mean function was compressed by halving the logarithmic sensitivities. The name of this disorder is potentially misleading because the comparable losses found across rod and cone vision suggest that the disorder is a generalized cone-rod dystrophy. Temporal acuity and temporal contrast sensitivity measures are broadly consistent with the defect in the voltage-gated potassium channel producing a nonlinear distortion of the photoreceptor response but after otherwise normal transduction processes.

Comparison of the cladding deformation measured during the Power Burst Facility loss-of-coolant accident in-pile experiments with recent Oak Ridge National Laboratory out-of-pile results

International Nuclear Information System (INIS)

Broughton, J.M.; McCardell, R.K.; MacDonald, P.E.

1981-01-01

A series of four large break loss-of-coolant accident fuel behavior experiments have been performed in the Power Burst Facility. The results of these experiments are briefly reviewed and compared with results from the ORNL multirod burst test program. The effect of cladding burst temperature and prior irradiation were investigated. The cladding strain of the previously irradiated test rods was more uniformly distributed around the cladding circumference and larger than for similar unirradiated test rods. The ORNL out-of-pile single rod test results are in good agreement with the Power Burst Facility (PBF) test results with unirradiated test rods, and the ORNL out-of-pile, single-rod test results with heated shrouds and the PBF test results with previously irradiated test rods are comparable

Some topics on safety analysis and accident nodalization of CAREM-25

International Nuclear Information System (INIS)

Gimenez, Marcelo O.; Zanocco, Pablo; Schlamp, Miguel A.; Ottaviani, Anahi; Garcia, Alicia

2000-01-01

The main goal of nuclear safety area in the CAREM Project Phase I, carried out during 1999, was to consolidate the safety systems design through an integral analysis of the reactor and the safety systems response to different accidental sequences. A primary circuit nodalization, including the steam generators, was done with RELAP5 code. The modeling of System 230 (absorber rods drive feed water system), System 1400 (purification and control volume system) and steam condensation on the absorber rods drive system and on RPV wall is implemented through boundary conditions. Also the Residual Heat Removal System and the Second Shutdown system are modeled. The reactor steady state at full power was calculated. The results agree quite well with design values. It can be said from the accident analysis that the nodalization responds properly. Further analysis should be done in order to qualify the nodalization and to compare benchmarks with other codes and experimental data. On the other hand, the steam dome model should be improved with more precise data about absorber rods drive system condensation, loss of heat and inner components layout. (author)

Control rod drive for vertical movement

International Nuclear Information System (INIS)

Suskov, I.I.; Gorjunov, V.S.; Zajcev, B.I.; Derevjankin, N.E.; Petrov, V.A.; Istomin, S.D.; Kovalencik, D.I.; Archipov, E.A.; Serebrjakov, V.I.; Kacalin, V.S.

1982-01-01

The control of the rod repositioning gear unit and the control unit of the profile grab of the control rod drive for the alkali metal-cooled fast breeder reactor is achieved by an electromotor being arranged outside the hermetic drive casing. The guide tube is directly repositioned by the rod repositioning gear unit. Coupling control of the drive with the control rod is done in the lower operative position of the control rod and that because of the interaction of the tie rod arranged on the spring-mounted control rod with the induction transmitter for the lower position of the control rod. In the transfer position the rod is fixed within the guide tube. (orig.)

Preliminary neutronic assessment for ATF (Accident Tolerant Fuel) based on iron alloy

International Nuclear Information System (INIS)

Abe, Alfredo; Carluccio, Thiago; Piovezan, Pamela; Giovedi, Claudia; Martins, Marcelo R.

2015-01-01

After Fukushima Daiichi nuclear accident in 2011, the nuclear fuel performance under accident condition became a very important issue and currently different research and development program are in progress toward to reliability and withstand under accident condition. These initiatives are known as ATF (Accident Tolerant Fuel) R and D program, which many countries with different research institutes, fuel vendors and others are nowadays involved. Accident Tolerant Fuel (ATF) can be defined as enhanced fuel which can tolerate loss of active cooling system capability for a considerably longer time period and the fuel/cladding system can be maintained without significant degradation and can also improve the fuel performance during normal operations and transients, as well as design-basis accident (DBA) and beyond design-basis (BDBA) accident. Different materials have being proposed as fuel cladding candidates considering thermo-mechanical properties and lower reaction kinetic with steam and slower hydrogen production. The aim of this work is to perform a neutronic assessment for several cladding candidates based on iron alloy considering a standard PWR fuel rod (fuel pellet and dimension). The purpose of the assessment is to address different parameters that might contribute for possible neutronic reactivity gain in order to overcome the penalty due to increase of neutron absorption in the cladding materials. All the neutronic assessment is performed using MCNP, Monte Carlo code. (author)

Some Examples of Accident Analyses for RB Reactor

International Nuclear Information System (INIS)

Pesic, M.

2002-01-01

The RB reactor is heavy water critical assembly operated in the Vinca Institute of Nuclear Sciences, Belgrade, Yugoslavia, since April 1959. The first Safety Analysis Report of the RB critical assembly was prepared in 1961/62. But, the first accidental analysis was done in late 1958 in aim the examine power transient and total equivalent doses received by the staff during the reactivity accident occurred on October 15, 1958. Since 1960, the RB reactor is modified few times. Beside initial natural uranium metal fuel rods, new fuel (TVR-S types) from 2% enriched metal uranium and 80% enriched UO 2 were available since 1962 and 1976, respectively. Also, modifications in control and safety systems of the reactor were done occasionally. Special reactor cores were created using all three types of fuel elements, among them, the coupled fast-thermal ones. Nuclear Safety Committee of the Vinca Institute, an independent regulatory body approved for usage all these modifications of the RB reactor. For those decisions of the Committee, the Preliminary Safety Analysis Reports were prepared that, beside proposed technical modifications and new regulation rules had included analyses of various possible accidents. Special attention is given and new methodology was proposed for thoroughly analyses of design based accidents related to coupled fast-thermal cores, that include reactor central zones filled by fuel elements without moderator. In these accidents, during assumed flooding of the fast zone by moderator, a very high reactivity could be inserted in the system with very high reactivity rate. It was necessary to provide that the safety system of the reactor had fast response to that accident and had enough high (negative) reactivity to shut down the reactor timely. In this paper, a brief overview of some accidents, methodology and computation tools used for the accident analyses at RB reactor are given. (author)

Out-of pile mechanical test: simulating reactivity initiated accident (RIA) of zircaloy-4 cladding tube

Energy Technology Data Exchange (ETDEWEB)

Lee, Myung Ho; Kim, Jun Hwan; Choi, Byoung Kwon; Jeong, Young Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2004-07-01

The ejection or drop of a control rod in a reactivity initiated accident (RIA) causes a sudden increase in reactor power and in turn deposits a large amount of energy into the fuel. In a RIA, cladding tubes bear thermal expansion due to sudden reactivity and may fail from the resulting mechanical damage. Thus, RIA can be one of the safety margin reducers because the oxide on the tubes makes their thickness to support the load less as well as hydrides from the corrosion reduce the ductility of the tubes. In a RIA, the peak of reactor power from reactivity change is about 0.1m second and the temperature of the cladding tubes increases up to 1000 .deg. C in several seconds. Although it is hard to fully simulate the situation, several attempts to measure the change of mechanical properties under a RIA situation has done using a reduction coil, ring tension tests with high speed. This research was done to see the effect of oxide on the change of circumferential strength and ductility of Zircaloy-4 tubes in a RIA. The ring stretch tensile tests were performed with the strain rate of 1/sec and 0.01/s to simulate a transient of the cladding tube under a RIA. Since the test results of the ring tensile test are very sensitive to the lubricant, the tests were also carried out to select a suitable lubricant before the test of oxided specimens.

Conservative performance analysis of a PWR nuclear fuel rod using the FRAPCON code

Energy Technology Data Exchange (ETDEWEB)

Oliveira, Fabio Branco Vaz de; Sabundjian, Gaiane, E-mail: fabio@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

In this paper, some of the preliminary results of the sensitivity and conservative analysis of a hypothetical pressurized water reactor fuel rod are presented, using the FRAPCON code as a basic and preparation tool for the future transient analysis, which will be carried out by the FRAPTRAN code. Emphasis is given to the evaluation of the cladding behavior, since it is one of the critical containment barriers of the fission products, generated during fuel irradiation. Sensitivity analyses were performed by the variation of the values of some parameters, which were mainly related with thermal cycle conditions, and taking into account an intermediate value between the realistic and conservative conditions for the linear heat generation rate parameter, given in literature. Time lengths were taken from typical nuclear power plant operational cycle, adjusted to the obtention of a chosen burnup. Curves of fuel and cladding temperatures, and also for their mechanical and oxidation behavior, as a function of the reactor operation's time, are presented for each one of the nodes considered, over the nuclear fuel rod. Analyzing the curves, it was possible to observe the influence of the thermal cycle on the fuel rod performance, in this preliminary step for the accident/transient analysis. (author)

Evidence linking coronal mass ejections with interplanetary magnetic clouds

International Nuclear Information System (INIS)

Wilson, R.M.; Hildner, E.

1983-12-01

Using proxy data for the occurrence of those mass ejections from the solar corona which are directed earthward, we investigate the association between the post-1970 interplanetary magnetic clouds of Klein and Burlaga and coronal mass ejections. The evidence linking magnetic clouds following shocks with coronal mass ejections is striking. Six of nine clouds observed at Earth were preceded an appropriate time earlier by meter-wave type II radio bursts indicative of coronal shock waves and coronal mass ejections occurring near central meridian. During the selected periods when no clouds were detected near Earth, the only type II bursts reported were associated with solar activity near the limbs. Where the proxy solar data to be sought are not so clearly suggested, that is, for clouds preceding interaction regions and clouds within cold magnetic enhancements, the evidence linking the clouds and coronal mass ejections is not as clear proxy data usually suggest many candidate mass-ejection events for each cloud. Overall, the data are consistent with and support the hypothesis suggested by Klein and Burlaga that magnetic clouds observed with spacecraft at 1 AU are manifestations of solar coronal mass ejection transients

Survey of potential light water reactor fuel rod failure mechanisms and damage limits

International Nuclear Information System (INIS)

Courtright, E.L.

1979-07-01

The findings and conclusions are presented of a survey to evaluate current information applicable to the development of fuel rod damage and failure limits for light water reactor fuel elements. The survey includes a review of past fuel failures, and identifies potential damage and failure mechanisms for both steady state operating conditions and postulated accident events. Possible relationships between the various damage and failure mechanisms are also proposed. The report identifies limiting criteria where possible, but concludes that sufficient data are not currently available in many important areas

Nuclear Energy Advanced Modeling and Simulation (NEAMS) Accident Tolerant Fuels High Impact Problem: Coordinate Multiscale U₃Si₂ Modeling

Energy Technology Data Exchange (ETDEWEB)

Gamble, K. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hales, J. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Miao, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Andersson, D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Zhang, Y. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-07-26

Since the events at the Fukushima-Daiichi nuclear power plant in March 2011 significant research has unfolded at national laboratories, universities and other institutions into alternative materials that have potential enhanced accident tolerance when compared to traditional \\uo~fuel zircaloy clad fuel rods. One of the potential replacement fuels is uranium silicide (\\usi) for its higher thermal conductivity and uranium density. The lower melting temperature is of potential concern during postulated accident conditions. Another disadvantage for \\usi~ is the lack of experimental data under power reactor conditions. Due to the aggressive development schedule for inserting some of the potential materials into lead test assemblies or rods by 2022~\\cite{bragg-sitton_2014} multiscale multiphysics modeling approaches have been used to provide insight into these materials. \\\\ \

Operational accidents in EL 2 and EL 3 between 1.1. 1957 and 1.7.1959 (1960)

International Nuclear Information System (INIS)

Balligand, P.

1960-01-01

The two most important accidents, costing 50 days out of operation for EL3 and 43 days for EL2, were due to the melting of a fuel rod through a cooling defect in a cell; the behaviour of the fuel could not otherwise be considered responsible. (author) [fr

Air pollution and heart failure: Relationship with the ejection fraction

Science.gov (United States)

Dominguez-Rodriguez, Alberto; Abreu-Afonso, Javier; Rodríguez, Sergio; Juarez-Prera, Ruben A; Arroyo-Ucar, Eduardo; Gonzalez, Yenny; Abreu-Gonzalez, Pedro; Avanzas, Pablo

2013-01-01

AIM: To study whether the concentrations of particulate matter in ambient air are associated with hospital admission due to heart failure in patients with heart failure with preserved ejection fraction and reduced ejection fraction. METHODS: We studied 353 consecutive patients admitted into a tertiary care hospital with a diagnosis of heart failure. Patients with ejection fraction of ≥ 45% were classified as having heart failure with preserved ejection fraction and those with an ejection fraction of < 45% were classified as having heart failure with reduced ejection fraction. We determined the average concentrations of different sizes of particulate matter (< 10, < 2.5, and < 1 μm) and the concentrations of gaseous pollutants (carbon monoxide, sulphur dioxide, nitrogen dioxide and ozone) from 1 d up to 7 d prior to admission. RESULTS: The heart failure with preserved ejection fraction population was exposed to higher nitrogen dioxide concentrations compared to the heart failure with reduced ejection fraction population (12.95 ± 8.22 μg/m3 vs 4.50 ± 2.34 μg/m3, P < 0.0001). Multivariate analysis showed that nitrogen dioxide was a significant predictor of heart failure with preserved ejection fraction (odds ratio ranging from (1.403, 95%CI: 1.003-2.007, P = 0.04) to (1.669, 95%CI: 1.043-2.671, P = 0.03). CONCLUSION: This study demonstrates that short-term nitrogen dioxide exposure is independently associated with admission in the heart failure with preserved ejection fraction population. PMID:23538391

Testing device for control rod drives

International Nuclear Information System (INIS)

Hayakawa, Toshifumi.

1992-01-01

A testing device for control rod drives comprises a logic measuring means for measuring an output signal from a control rod drive logic generation circuit, a control means for judging the operation state of a control rod and a man machine interface means for outputting the result of the judgement. A driving instruction outputted from the control rod operation device is always monitored by the control means, and if the operation instruction is stopped, a testing signal is outputted to the control rod control device to simulate a control rod operation. In this case, the output signal of the control rod drive logic generation circuit is held in a control rod drive memory means and intaken into a logic analysis means for measurement and an abnormality is judged by the control means. The stopping of the control rod drive instruction is monitored and the operation abnormality of the control rod is judged, to mitigate the burden of an operator. Further, the operation of the control rod drive logic generation circuit can be confirmed even during a nuclear plant operation by holding the control rod drive instruction thereby enabling to improve maintenance efficiency. (N.H.)

Characteristics of debris in the lower head of a BWR in different severe accident scenarios

International Nuclear Information System (INIS)

Phung, Viet-Anh; Galushin, Sergey; Raub, Sebastian; Goronovski, Andrei; Villanueva, Walter; Kööp, Kaspar; Grishchenko, Dmitry; Kudinov, Pavel

2016-01-01

Highlights: • Station blackout scenario with delayed recovery of safety systems in a Nordic BWR is considered. • Genetic algorithm and random sampling methods are used to explore accident scenario domain. • Main groups of scenarios are identified. • Ranges and distributions of characteristics of debris bed in the lower head are determined. - Abstract: Nordic boiling water reactors (BWRs) adopt ex-vessel debris cooling to terminate severe accident progression. Core melt released from the vessel into a deep pool of water is expected to fragment and form a coolable debris bed. Characteristics of corium melt ejection from the vessel determine conditions for molten fuel–coolant interactions (FCI) and debris bed formation. Non-coolable debris bed or steam explosion can threaten containment integrity. Vessel failure and melt ejection mode are determined by the in-vessel accident progression. Characteristics (such as mass, composition, thermal properties, timing of relocation, and decay heat) of the debris bed formed in the process of core relocation into the vessel lower plenum define conditions for the debris reheating, remelting, melt-vessel structure interactions, vessel failure and melt release. Thus core degradation and relocation are important sources of uncertainty for the success of the ex-vessel accident mitigation strategy. The goal of this work is improve understanding how accident scenario parameters, such as timing of failure and recovery of different safety systems can affect characteristics of the debris in the lower plenum. Station blackout scenario with delayed power recovery in a Nordic BWR is considered using MELCOR code. The recovery timing and capacity of safety systems were varied using genetic algorithm (GA) and random sampling methods to identify two main groups of scenarios: with relatively small ( 100 tons) amount of relocated debris. The domains are separated by the transition regions, in which relatively small variations of the input

Characteristics of debris in the lower head of a BWR in different severe accident scenarios

Energy Technology Data Exchange (ETDEWEB)

Phung, Viet-Anh, E-mail: vaphung@kth.se; Galushin, Sergey, E-mail: galushin@kth.se; Raub, Sebastian, E-mail: raub@kth.se; Goronovski, Andrei, E-mail: andreig@kth.se; Villanueva, Walter, E-mail: walterv@kth.se; Kööp, Kaspar, E-mail: kaspar@safety.sci.kth.se; Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se

2016-08-15

Highlights: • Station blackout scenario with delayed recovery of safety systems in a Nordic BWR is considered. • Genetic algorithm and random sampling methods are used to explore accident scenario domain. • Main groups of scenarios are identified. • Ranges and distributions of characteristics of debris bed in the lower head are determined. - Abstract: Nordic boiling water reactors (BWRs) adopt ex-vessel debris cooling to terminate severe accident progression. Core melt released from the vessel into a deep pool of water is expected to fragment and form a coolable debris bed. Characteristics of corium melt ejection from the vessel determine conditions for molten fuel–coolant interactions (FCI) and debris bed formation. Non-coolable debris bed or steam explosion can threaten containment integrity. Vessel failure and melt ejection mode are determined by the in-vessel accident progression. Characteristics (such as mass, composition, thermal properties, timing of relocation, and decay heat) of the debris bed formed in the process of core relocation into the vessel lower plenum define conditions for the debris reheating, remelting, melt-vessel structure interactions, vessel failure and melt release. Thus core degradation and relocation are important sources of uncertainty for the success of the ex-vessel accident mitigation strategy. The goal of this work is improve understanding how accident scenario parameters, such as timing of failure and recovery of different safety systems can affect characteristics of the debris in the lower plenum. Station blackout scenario with delayed power recovery in a Nordic BWR is considered using MELCOR code. The recovery timing and capacity of safety systems were varied using genetic algorithm (GA) and random sampling methods to identify two main groups of scenarios: with relatively small (<20 tons) and large (>100 tons) amount of relocated debris. The domains are separated by the transition regions, in which relatively small

Sensitivity analysis of local uncertainties in large break loss-of-coolant accident (LB-LOCA) thermo-mechanical simulations

Energy Technology Data Exchange (ETDEWEB)

Arkoma, Asko, E-mail: asko.arkoma@vtt.fi; Ikonen, Timo

2016-08-15

Highlights: • A sensitivity analysis using the data from EPR LB-LOCA simulations is done. • A procedure to analyze such complex data is outlined. • Both visual and quantitative methods are used. • Input factors related to core design are identified as most significant. - Abstract: In this paper, a sensitivity analysis for the data originating from a large break loss-of-coolant accident (LB-LOCA) analysis of an EPR-type nuclear power plant is presented. In the preceding LOCA analysis, the number of failing fuel rods in the accident was established (Arkoma et al., 2015). However, the underlying causes for rod failures were not addressed. It is essential to bring out which input parameters and boundary conditions have significance to the outcome of the analysis, i.e. the ballooning and burst of the rods. Due to complexity of the existing data, the first part of the analysis consists of defining the relevant input parameters for the sensitivity analysis. Then, selected sensitivity measures are calculated between the chosen input and output parameters. The ultimate goal is to develop a systematic procedure for the sensitivity analysis of statistical LOCA simulation that takes into account the various sources of uncertainties in the calculation chain. In the current analysis, the most relevant parameters with respect to the cladding integrity are the decay heat power during the transient, the thermal hydraulic conditions in the rod’s location in reactor, and the steady-state irradiation history of the rod. Meanwhile, the tolerances in fuel manufacturing parameters were found to have negligible effect on cladding deformation.

Safety against releases in severe accidents. Final report

Energy Technology Data Exchange (ETDEWEB)

Lindholm, I.; Berg, Oe.; Nonboel, E. [eds.

1997-12-01

The work scope of the RAK-2 project has involved research on quantification of the effects of selected severe accident phenomena for Nordic nuclear power plants, development and testing of a computerised accident management support system and data collection and description of various mobile reactors and of different reactor types existing in the UK. The investigations of severe accident phenomena focused mainly on in-vessel melt progression, covering a numerical assessment of coolability of a degraded BWR core, the possibility and consequences of a BWR reactor to become critical during reflooding and the core melt behavior in the reactor vessel lower plenum. Simulant experiments were carried out to investigate lower head hole ablation induced by debris discharge. In addition to the in-vessel phenomena, a limited study on containment response to high pressure melt ejection in a BWR and a comparative study on fission product source term behaviour in a Swedish PWR were performed. An existing computerised accident management support system (CAMS) was further developed in the area of tracking and predictive simulation, signal validation, state identification and user interface. The first version of a probabilistic safety analysis module was developed and implemented in the system. CAMS was tested in practice with Barsebaeck data in a safety exercise with the Swedish nuclear authority. The descriptions of the key features of British reactor types, AGR, Magnox, FBR and PWR were published as data reports. Separate reports were issued also on accidents in nuclear ships and on description of key features of satellite reactors. The collected data were implemented in a common Nordic database. (au) 39 refs.

Polarized DNA Ejection from the Herpesvirus Capsid

Science.gov (United States)

Newcomb, William W.; Cockrell, Shelley K.; Homa, Fred L.; Brown, Jay C.

2009-01-01

Ejection of DNA from the capsid is an early step in infection by all herpesviruses. Ejection or DNA uncoating occurs after a parental capsid has entered the host cell cytoplasm, migrated to the nucleus and bound to a nuclear pore. DNA exits the capsid through the portal vertex and proceeds by way of the nuclear pore complex into the nucleoplasm where it is transcribed and replicated. Here we describe use of an in vitro uncoating system to determine which genome end exits first from the herpes simplex virus (HSV-1) capsid. Purified DNA-containing capsids were bound to a solid surface and warmed under conditions in which some, but not all, of the DNA was ejected. Restriction endonuclease digestion was then used to identify the genomic origin of the ejected DNA. The results support the view that the S segment end exits the capsid first. Preferential release at the S end demonstrates that herpesvirus DNA uncoating conforms to the paradigm in dsDNA bacteriophage where the last end packaged is the first to be ejected. Release of HSV-1 DNA beginning at the S end causes the first gene to enter the host cell nucleus to be α4, a transcription factor required for expression of early genes. PMID:19631662

Failed fuel rod detector

Energy Technology Data Exchange (ETDEWEB)

Uchida, Katsuya; Matsuda, Yasuhiko

1984-05-02

The purpose of the project is to enable failed fuel rod detection simply with no requirement for dismantling the fuel assembly. A gamma-ray detection section is arranged so as to attend on the optional fuel rods in the fuel assembly. The fuel assembly is adapted such that a gamma-ray shielding plate is detachably inserted into optional gaps of the fuel rods or, alternatively, the fuel assembly can detachably be inserted to the gamma-ray shielding plate. In this way, amount of gaseous fission products accumulated in all of the plenum portions in the fuel rods as the object of the measurement can be determined without dismantling the fuel assembly. Accordingly, by comparing the amounts of the gaseous fission products, the failed fuel rod can be detected.

RodPilotR - The Innovative and Cost-Effective Digital Control Rod Drive Control System for PWRs

International Nuclear Information System (INIS)

Baron, Clemens

2008-01-01

With RodPilot, AREVA NP offers an innovative and cost-effective system for controlling control rods in Pressurized Water Reactors. RodPilot controls the three operating coils of the control rod drive mechanism (lift, moveable gripper and stationary gripper coil). The rods are inserted into or withdrawn from the core as required by the Reactor Control System. The system combines modern components, state-of-the-art logic and a proven electronic control rod drive control principle to provide enhanced reliability and lower maintenance costs. (author)

Integrated control rod monitoring device

International Nuclear Information System (INIS)

Saito, Katsuhiro

1997-01-01

The present invention provides a device in which an entire control rod driving time measuring device and a control rod position support device in a reactor building and a central control chamber are integrated systematically to save hardwares such as a signal input/output device and signal cables between boards. Namely, (1) functions of the entire control rod driving time measuring device for monitoring control rods which control the reactor power and a control rod position indication device are integrated into one identical system. Then, the entire devices can be made compact by the integration of the functions. (2) The functions of the entire control rod driving time measuring device and the control rod position indication device are integrated in a central operation board and a board in the site. Then, the place for the installation of them can be used in common in any of the cases. (3) The functions of the entire control rod driving time measuring device and the control rod position indication device are integrated to one identical system to save hardware to be used. Then, signal input/output devices and drift branching panel boards in the site and the central operation board can be saved, and cables for connecting both of the boards is no more necessary. (I.S.)

Rod drive and latching mechanism

International Nuclear Information System (INIS)

Veronesi, L.; Sherwood, D.G.

1982-01-01

Hydraulic drive and latching mechanisms for driving reactivity control mechanisms in nuclear reactors are described. Preferably, the pressurized reactor coolant is utilized to raise the drive rod into contact with and to pivot the latching mechanism so as to allow the drive rod to pass the latching mechanism. The pressure in the housing may then be equalized which allows the drive rod to move downwardly into contact with the latching mechanism but to hold the shaft in a raised position with respect to the reactor core. Once again, the reactor coolant pressure may be utilized to raise the drive rod and thus pivot the latching mechanism so that the drive rod passes above the latching mechanism. Again, the mechanism pressure can be equalized which allows the drive rod to fall and pass by the latching mechanism so that the drive rod approaches the reactor core. (author)

A study of the large break loss-of-coolant accident in the Angra-1 nuclear power plant

International Nuclear Information System (INIS)

Borges, E.M.

1984-01-01

The simulation of the Angra-I nuclear power plant under the condition of large break loss of coolant accident is presented, the thermal-hydraulic analysis of the primary circuit during each phase of the acident and thermal analysis of the hottest fuel rod curing reflooding are shown. Computer codes RELAP4/MOD5 (options EM and FLOOD) and TOODEE 2 are used to perform these computations. Fuel rod peak temperatures reached during the simulation are below the permissible levels. However, during the reflooding phase; the maximum oxidation of the cladding exceeds the limit of 0.17 times the original cladding thickness. (Author) [pt

Control rods

International Nuclear Information System (INIS)

Koga, Isao; Masuoka, Ryuzo.

1979-01-01

Purpose: To prevent fuel element failures during power conditioning by removing liquid absorbents in poison tubes of control rods in a fast power up step and extracting control rods to slightly increase power in a medium power up step. Constitution: A plurality of poison tubes are disposed in a coaxial or plate-like arrangement and divided into a region capable of compensating the reactivity from the initial state at low temperature to 40% power operation and a region capable of compensating the reactivity in the power up above 40% power operation. Soluble poisons are used as absorbers in the poison tubes corresponding to above 40% power operation and they are adapted to be removed independently from the driving of control rods. The poison tubes filled with the soluble absorbers are responsible for the changes in the reactivity from the initial state at low temperature to the medium power region and the reactivity control is conducted by the elimination of liquid absorbers from the poison tubes. In the succeeding slight power up region above the medium power, power up is proceeding by extracting the control rods having remaining poison tubes filled with solid or liquid absorbers. (Seki, T.)

PBF/LOFT Lead Rod Test Program experiment operating specification

International Nuclear Information System (INIS)

Varacalle, D.J. Jr.

1978-11-01

The PBF/LOFT Lead Rod (LLR) Test Program is being conducted to provide experimental information on the behavior of nuclear fuel under normal and accident conditions in the Power Burst Facility at the Idaho National Engineering Laboratory. Understanding the behavior of light-water reactors (LWR) under loss-of-coolant conditions is a major objective of the NRC Reactor Safety Research Program. The Loss of Fluid Test (LOFT) facility is the major testing facility to evaluate the systems response of an LWR over a wide range of Loss of Coolant Experment (LOCE) conditions. As such, the LOFT core is intended to be used for sequential LOCE tests provided no significant fuel rod failures occur. The PFB/LLR tests are designed to simulate the test conditions for the LOFT Power Ascension Tests L2-2 through L2-5. The test program has been designed to provide a parametric evaluation of the LOFT fuel over a wide range of power. Thus, a relatively accurate assessment of the state of the LOFT core after the completion of each subtest and the anticipated effect of the next test can be obtained by utilizing a combination of LLR test data and analytical predictions. Specifications for the test program are presented

Review and compilation of criticality accidents in nuclear fuel processing facilities outside of Japan

International Nuclear Information System (INIS)

Watanabe, Norio; Tamaki, Hitoshi

2000-04-01

On September 30, 1999, a criticality accident occurred at the Tokai-mura uranium processing plant operated by JCO Co., Ltd., which resulted in the first nuclear accident involving a fatality, in Japan, and forced the residents in the vicinity of the site to be evacuated and be sheltered indoors. There have now been 21 criticality accidents reported in nuclear fuel processing facilities in foreign countries: seven in the United States, one in the United Kingdom and thirteen in Russia. Most of them occurred during the period from mid-1950's to mid-1960's, but one criticality accident tool place in Russian in 1997. This report reviews and compiles the published information on these accidents, including the latest information, focusing on the event sequence, the consequence of accident, and the cause of accident. The observations from the reviews are summarized as follows: Twenty of the 21 accidents occurred with the fissile material in a liquid. Twenty of the 21 accidents occurred in vessels/tanks with unfavorable geometry but one occurred in the vessel with favorable geometry. There were seven fatalities that were involved in five accidents. Three accidents involved a re-criticality condition caused by inadequate operator actions and two of them led to the death of the operators. One accident reached a re-criticality condition several hours after the first excursion was terminated by injecting borated water into the affected vessel. This accident implies the possibility that the borated water injection might not be effective to the criticality termination due to solubility of boric acid. Mechanisms of the criticality termination vary as follows: ejection or splashing of the solution at the time of power excursion, boiling or evaporation, addition of neutron poisons, or manual draining of solutions. (author)

Review and compilation of criticality accidents in nuclear fuel processing facilities outside of Japan

Energy Technology Data Exchange (ETDEWEB)

Watanabe, Norio [Planning and Analysis Division, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Tamaki, Hitoshi [Department of Safety Research Technical Support, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)

2000-04-01

On September 30, 1999, a criticality accident occurred at the Tokai-mura uranium processing plant operated by JCO Co., Ltd., which resulted in the first nuclear accident involving a fatality, in Japan, and forced the residents in the vicinity of the site to be evacuated and be sheltered indoors. There have now been 21 criticality accidents reported in nuclear fuel processing facilities in foreign countries: seven in the United States, one in the United Kingdom and thirteen in Russia. Most of them occurred during the period from mid-1950's to mid-1960's, but one criticality accident tool place in Russian in 1997. This report reviews and compiles the published information on these accidents, including the latest information, focusing on the event sequence, the consequence of accident, and the cause of accident. The observations from the reviews are summarized as follows: Twenty of the 21 accidents occurred with the fissile material in a liquid. Twenty of the 21 accidents occurred in vessels/tanks with unfavorable geometry but one occurred in the vessel with favorable geometry. There were seven fatalities that were involved in five accidents. Three accidents involved a re-criticality condition caused by inadequate operator actions and two of them led to the death of the operators. One accident reached a re-criticality condition several hours after the first excursion was terminated by injecting borated water into the affected vessel. This accident implies the possibility that the borated water injection might not be effective to the criticality termination due to solubility of boric acid. Mechanisms of the criticality termination vary as follows: ejection or splashing of the solution at the time of power excursion, boiling or evaporation, addition of neutron poisons, or manual draining of solutions. (author)

Cleanup of large areas contaminated as a result of a nuclear accident

International Nuclear Information System (INIS)

1989-01-01

The purposes of the report are to provide an overview of the methodology and technology available to clean up contaminated areas and to give preliminary guidance on matters related to the planning, implementation and management of such cleanups. This report provides an integrated overview of important aspects related to the cleanup of very large areas contaminated as a result of a serious nuclear accident, including information on methods and equipment available to: characterize the affected area and the radioactive fallout; stabilize or isolate the contamination; and clean up contaminated urban, rural and forested areas. The report also includes brief sections on planning and management considerations and the transport and disposal of the large volumes of wastes arising from such cleanups. For the purposes of this report, nuclear accidents which could result in the deposition of decontamination over large areas if the outer containment fails badly include: 1) An accident with a nuclear weapon involving detonation of the chemical high explosive but little, if any, nuclear fission. 2) A major loss of medium/high level liquid waste (HLLW) due to an explosion/fire at a storage site for such waste. 3) An accident at a nuclear power plant (NPP), for example a loss of coolant accident, which results in some core disruption and fuel melting. 4) An accident at an NPP involving an uncontrolled reactivity excursion resulting in the violent ejection of a reactor core material and rupture of the containment building. 117 refs, 32 figs, 12 tabs

Fuel rod technology

International Nuclear Information System (INIS)

Bezold, H.; Romeiser, H.J.

1979-07-01

By extensive mechanization and automation of the fuel rod production, also at increasing production numbers, an efficient production shall be secured, simultaneously corresponding to the high quality standard of the fuel rods. The works done up to now concentrated on the lay out of a rough concept for a mechanized production course. Detail-studies were made for the problems of fuel rod humidity, filling and resistance welding. Further promotion of this project and thus further report will be stopped, since the main point of these works is the production technique. (orig.) [de

Accidents - Chernobyl accident; Accidents - accident de Tchernobyl

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

This file is devoted to the Chernobyl accident. It is divided in four parts. The first part concerns the accident itself and its technical management. The second part is relative to the radiation doses and the different contaminations. The third part reports the sanitary effects, the determinists ones and the stochastic ones. The fourth and last part relates the consequences for the other European countries with the case of France. Through the different parts a point is tackled with the measures taken after the accident by the other countries to manage an accident, the cooperation between the different countries and the groups of research and studies about the reactors safety, and also with the international medical cooperation, specially for the children, everything in relation with the Chernobyl accident. (N.C.)

Proceedings of the Second OECD/NEA Organisation Meeting on Increased Accident Tolerance of Fuels for LWRs

International Nuclear Information System (INIS)

Massara, Simone; ); Bragg-Sitton, Shannon; Braase, Lori; Merrill, Brad; Teague, Melissa; Stanek, Chris R.; Montgomery, Robert H.; Ott, Larry J.; Robb, Kevin; Snead, Lance; Farmer, Mitch; Billone, Michael C.; Todosow, Michael; Brown, Nicholas; Brachet, J.C.; Le Flem, M.; Sauder, C.; Idarraga-Trujillo, I.; Michaux, A.; Lorrette, C.; Le Saux, M.; Blanpain, P.; Park, Jeong-Yong; Yang, Jae-Ho; Kim, Weon-Ju; Koo, Yang-Hyun; Liu, T.; Hallstadius, Lars; Lahoda, Ed; Waeckel, N.; Bonnet, J.M.; Vitanza, Carlo; Ohta, Hirokazu; Ogata, Takanari; Nakamura, Kinya; Dyck, Gary; Inozemtsev, Victor; )

2013-01-01

Under the guidance of the OECD-NEA Nuclear Science Committee, the expert group acts as a forum for scientific and technical information exchange on advanced light water reactor (LWR) fuels with enhanced accident tolerance. The expert group focusses on the fundamental properties and behaviour under normal operations and accident conditions for advanced core materials and components (fuels, cladding, control rods, etc.). The materials considered are applicable to Gen II and Gen III Light Water Reactors, as well as Gen III+ reactors under construction. The objective of the expert group is to define and coordinate a programme of work to help advance the scientific knowledge needed to provide the technical underpinning for the development of advanced LWR fuels with enhanced accident tolerance compared to currently used zircaloy/UO 2 fuel systems, as well as other non-fuel core components with important roles in LWR performance under accident conditions. This document brings together the available presentations (slides) given at the 2. Meeting on Increased Accident Tolerance of Fuels for LWRs. Content: 1 - Overview of the exchanges after the December-2012 Workshop through the discussion forum established at the OECD-NEA (S. Massara, NEA); 2 - Metrics Development for Enhanced Accident Tolerant LWR Fuels (S. Bragg-Sitton, INL); 3 - Candidate ATF Clad Technologies and Key Feasibility Issues (L. Snead, ORNL); 4 - CEA studies on nuclear fuel claddings for LWRs enhanced accident tolerant fuel. Some recent results, pending issues and prospects (J.C. Brachet, CEA); 5 - Current status on the accident tolerant fuel development in the Republic of Korea (J.Y. Park, J.H. Chang, KAERI); 6 - The current status of fuel R and D in the P.R. of China (T. Liu, CGN). Session 2: Key elements for a work programme on ATF: 7 - Beneficial characteristics of ATF (metrics) (L. Hallstadius, Westinghouse); 8 - Reactor types of interest (applicability) (L. Ott, ORNL); 9 - Impact on normal operations

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

International Nuclear Information System (INIS)

Gomes, Daniel de Souza

2014-01-01

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)

Digital control rod blocking monitor

International Nuclear Information System (INIS)

Funayama, Yoshio.

1996-01-01

The present invention system is used for monitoring of a power region of a reactor, and used for monitoring of simultaneous withdrawal of a plurality of control rods without increasing the size or complicating the system. Namely, the system processes signals from a neutron flux detectors at the periphery of control rods controlled for withdrawal. As a result of the processing, the digital monitoring system generates an alarm when the reactor power at the periphery of the control rods fluctuates exceeding an allowable range. In the system, a control rod information forming means prepares frame data comprising front data, positions of the control rods to be withdrawn, frame numbers and completion data. A serial data transmitting means transmits the frame data successively as repeating frame data rows. A control rod information receiving means takes up the frame data of each of control rods to be withdrawn from the transmitted frame data rows. Since the system of the present invention can monitor the withdrawal of a plurality of control rods simultaneously without increasing the size or complicating the system, cost can be saved and the maintenance can be improved. (I.S.)

Analysis of Severe Accident for the SFP under the Condition of Drainage using MELCOR

Energy Technology Data Exchange (ETDEWEB)

Oh, Jung-Min; Pack, Jae-Woo [Jeju National University, Jeju (Korea, Republic of)

2015-10-15

This study aims to analyze the effect of a LOCA of the spent fuel pool. We use the MECORE 1.8.6 code to compute the variation of the fuel cladding temperature after a completer loss of the cooling water in the spent fuel pool. A loss of coolant accident in a typical spent fuel pool has been simulated using the MELCOR 1.8.6 code to see the variation of key parameters such as the oxygen concentration in the fuel assembly region and the cladding temperature. In a commercial nuclear power plant, highly radioactive spent fuel assemblies unloaded from the nuclear reactor core are typically stored for a period of time in the spent fuel pool to reduce the radioactivity. The spent fuel assemblies are usually placed in long square racks. It is known that in the progress of the Fukushima nuclear power plant accident, the cooling water in the spent fuel storage was completely lost and the fuel was heated up and damaged. The simulation result shows that the cladding temperature exceeds the rupture temperature in most of the fuel rods and some part of the fuel rods suffers melting of the cladding.

The BWR Hybrid 4 control rod

International Nuclear Information System (INIS)

Gross, H.; Fuchs, H.P.; Lippert, H.J.; Dambietz, W.

1988-01-01

The service life of BWR control rods designed in the past has been unsatisfactory. The main reason was irradiation assisted stress corrosion cracking of B 4 C rods caused by external swelling of the B 4 C powder. By this reason KWU developed an improved BWR control rod (Hybrid 4 control rod) with extended service life and increased control rod worth. It also allows the procedure for replacing and rearranging fuel assemblies to be considerably simplified. A complete set of Hydbrid 4 control rods is expected to last throughout the service life of a plant (assumption: ca. 40 years) if an appropriate control rod reshuffling management program is used. (orig.)

Application of the SCANAIR code for VVER RIA conditions - Boron dilution accident

International Nuclear Information System (INIS)

Arffman, A.; Cazalis, B.

2010-01-01

This paper consists of two parts. In part A, RIA pulse tests conducted at the Russian BIGR reactor are being analysed at IRSN with SCANAIR V6 fuel performance code as a part of the code validation for VVER fuel. Recently a new version of the SCANAIR code was made available to VTT Technical Research Centre of Finland, and part B of the paper covers the introduction of the code version at VTT by a calculation of a hypothetical boron dilution accident in a VVER-440 power reactor. Concerning part A, it appears that the SCANAIR V6 version, including a BIGR/NSRR heat transfer model, validated by Japanese NSRR experiments, and a Norton viscoplastic clad mechanical behaviour, is able to simulate the rod thermal behaviour in BIGR tests. Concerning the clad mechanics, it has been seen that a pellet swelling model is able to simulate the average rod deformation. Nonetheless, the current clad creep model associated with the free volume equilibrium assumption is not suited to predict the maximum clad deformation and the possible post DNB rod failure because they do not simulate local balloons. Furthermore, it has been shown that the clad deformation is strongly dependent on transient gas transfer. Concerning part B, a boron dilution accident previously calculated with SCANAIR V2 was recalculated with SCANAIR V6. A limited amount of result parameters were compared with the results of VTT's neutronics code TRAB. Divergence problems encountered previously when reaching the DNB limit were not present anymore. Fuel and cladding temperatures produced by SCANAIR were in good agreement with those calculated with TRAB

Coronal Mass Ejections

CERN Document Server

Kunow, H; Linker, J. A; Schwenn, R; Steiger, R

2006-01-01

It is well known that the Sun gravitationally controls the orbits of planets and minor bodies. Much less known, however, is the domain of plasma fields and charged particles in which the Sun governs a heliosphere out to a distance of about 15 billion kilometers. What forces activates the Sun to maintain this power? Coronal Mass Ejections (CMEs) and their descendants are the troops serving the Sun during high solar activity periods. This volume offers a comprehensive and integrated overview of our present knowledge and understanding of Coronal Mass Ejections (CMEs) and their descendants, Interplanetary CMEs (ICMEs). It results from a series of workshops held between 2000 and 2004. An international team of about sixty experimenters involved e.g. in the SOHO, ULYSSES, VOYAGER, PIONEER, HELIOS, WIND, IMP, and ACE missions, ground observers, and theoreticians worked jointly on interpreting the observations and developing new models for CME initiations, development, and interplanetary propagation. The book provides...

Burnable poison rod

International Nuclear Information System (INIS)

Natsume, Tomohiro.

1988-01-01

Purpose: To decrease the effect of water elimination and the effect of burn-up residue boron, thereby reduce the effect of burnable poison rods as the neutron poisons at the final stage of reactor core lifetime. Constitution: In a burnable poison rod according to the present invention, a hollow burnable poison material is filled in an external fuel can, an inner fuel can mounted with a carbon rod is inserted to the hollow portion of the burnable poison material and helium gases are charged in the outer fuel can. In such a burnable poison rod, the reactivity worths after the burning are reduced to one-half as compared with the conventional case. Accordingly, since the effect of the burnable poison as the neutron poisons is reduced at the final stage of the reactor core of lifetime, the excess reactivity of the reactor core is increased. (Horiuchi, T.)

Calculations of combined radiation and convection heat transfer in rod bundles under emergency cooling conditions

International Nuclear Information System (INIS)

Sun, K.H.; Gonzalez-Santalo, J.M.; Tien, C.L.

1976-01-01

A model has been developed to calculate the heat transfer coefficients from the fuel rods to the steam-droplet mixture typical of Boiling Water Reactors under Emergency Core Cooling System (ECCS) operation conditions during a postulated loss-of-coolant accident. The model includes the heat transfer by convection to the vapor, the radiation from the surfaces to both the water droplets and the vapor, and the effects of droplet evaporation. The combined convection and radiation heat transfer coefficient can be evaluated with respect to the characteristic droplet size. Calculations of the heat transfer coefficient based on the droplet sizes obtained from the existing literature are consistent with those determined empirically from the Full-Length-Emergency-Cooling-Heat-Transfer (FLECHT) program. The present model can also be used to assess the effects of geometrical distortions (or deviations from nominal dimensions) on the heat transfer to the cooling medium in a rod bundle

Response of unirradiated and irradiated PWR fuel rods tested under power-cooling-mismatch conditions

International Nuclear Information System (INIS)

MacDonald, P.E.; Quapp, W.J.; Martinson, Z.R.; McCardell, R.K.; Mehner, A.S.

1978-01-01

This report summarizes the results from the single-rod power-cooling-mismatch (PCM) and irradiation effects (IE) tests conducted to date in the Power Burst Facility (PBF) at the U.S. DOE Idaho National Engineering Laboratory. This work was performed for the U.S. NRC under contact to the Department of Energy. These tests are part of the NRC Fuel Behavior Program, which is designed to provide data for the development and verification of analytical fuel behavior models that are used to predict fuel response to abnormal or postulated accident conditions in commercial LWRs. The mechanical, chemical and thermal response of both previously unirradiated and previously irradiated LWR-type fuel rods tested under power-cooling-mismatch condition is discussed. A brief description of the test designs is presented. The results of the PCM thermal-hydraulic studies are summarized. Primary emphasis is placed on the behavior of the fuel and cladding during and after stable film boiling. (orig.) [de

Coronal mass ejections and coronal structures

International Nuclear Information System (INIS)

Hildner, E.; Bassi, J.; Bougeret, J.L.

1986-01-01

Research on coronal mass ejections (CMF) took a variety of forms, both observational and theoretical. On the observational side there were: case studies of individual events, in which it was attempted to provide the most complete descriptions possible, using correlative observations in diverse wavelengths; statistical studies of the properties of CMEs and their associated activity; observations which may tell us about the initiation of mass ejections; interplanetary observations of associated shocks and energetic particles; observations of CMEs traversing interplanetary space; and the beautiful synoptic charts which show to what degree mass ejections affect the background corona and how rapidly (if at all) the corona recovers its pre-disturbance form. These efforts are described in capsule form with an emphasis on presenting pictures, graphs, and tables so that the reader can form a personal appreciation of the work and its results

WASA-BOSS. Development and application of Severe Accident Codes. Evaluation and optimization of accident management measures. Subproject F. Contributions to code validation using BWR data and to evaluation and optimization of accident management measures. Final report

International Nuclear Information System (INIS)

Di Marcello, Valentino; Imke, Uwe; Sanchez Espinoza, Victor

2016-09-01

The exact knowledge of the transient course of events and of the dominating processes during a severe accident in a nuclear power station is a mandatory requirement to elaborate strategies and measures to minimize the radiological consequences of core melt. Two typical experiments using boiling water reactor assemblies were modelled and simulated with the severe accident simulation code ATHLET-CD. The experiments are related to the early phase of core degradation in a boiling water reactor. The results reproduce the thermal behavior and the hydrogen production due to oxidation inside the bundle until relocation of material by melting. During flooding of the overheated assembly temperatures and hydrogen oxidation are under estimated. The deviations from the experimental results can be explained by the missing model to simulate bore carbide oxidation of the control rods. On basis of a hypothetical loss of coolant accident in a typical German boiling water reactor the effectivity of flooding the partial degraded core is investigated. This measure of mitigation is efficient and prevents failure of the reactor pressure vessel if it starts before molten material is relocated into the lower plenum. Considerable amount of hydrogen is produced by oxidation of the metallic components.

Numerical simulation of temperature's sensitivity of chamfer hole's resistance on hydraulic step cylinder

International Nuclear Information System (INIS)

Jinhua, Wang; Hanliang, Bo; Wenxiang, Zheng; Jinnong, Yang

2003-01-01

The control rod drive is a very important device for controlling nuclear reactor startup, operation, shut down, and power change. The ability of the control rod drive to move safely and reliably directly relates to reactor safety. The Hydraulic Control Rod Drive System (HCRDS) is a new type of control rod drive system developed by the Institute of Nuclear Energy Technology (INET) of Tsinghua University for Nuclear Heating Reactors. The HCRDS, designed using the hydrodynamic principle, has many advantages, including having the structure complete in the vessel, no possible ejection accident, short drive line, simple movable parts structure and safe shutdown during accidents. The hydraulic step cylinder is the key part for the HCRDS. In the process of reactor startup, the variation of temperature could make the water's density and viscosity change, and the force from the water flow would change accordingly. These factors could influence the performance of the hydraulic step cylinder. In this paper, the temperature sensitivity of the chamfer hole's resistance in the hydraulic step cylinder was studied with the Computational Fluid Dynamics (CFD) program CFX5.5. The results were satisfactory: the discipline of variation of the chamfer hole's resistance with the outer tube's position was the same at different temperatures, the discrepancy of the chamfer hole's resistance was small for the same position at different temperatures, the chamfer hole's resistance decreased gradually with the increase of temperature, and the decrease extent was relatively small

AgInCd control rod failure in the QUENCH-13 bundle test

International Nuclear Information System (INIS)

Sepold, L.; Lind, T.; Csordas, A. Pinter; Stegmaier, U.; Steinbrueck, M.; Stuckert, J.

2009-01-01

during the pre-reflood phases). Posttest examinations of bundle structures revealed the presence of only little relocated AgInCd melt in the form of rivulets, mainly in the coolant channels surrounding the control rod simulator and at axial elevations between the third (0.55 m) and first spacer grids (-0.1 m). Results of QUENCH-13 on the onset of absorber rod failure are in agreement with CORA results of nine experiments each containing one or more AgInCd/stainless steel/Zircaloy-4 control rod assemblies. Bundle degradation triggered by early melt formation was, however, more pronounced in the CORA experiments with maximum bundle temperatures of ∼2300 K (compared to ∼1800 K in QUENCH-13). Consequently, QUENCH-13 allowed studying the initiation of absorber rod failure by eutectic reactions of SS-Zr, and later on of AgInCd-Zr, as well as the redistribution of the absorber material within the test bundle. Furthermore, input data for modeling of aerosol release during severe accidents are considered as benefits of the experiment.

Biomechanics of lumbar cortical screw-rod fixation versus pedicle screw-rod fixation with and without interbody support.

Science.gov (United States)

Perez-Orribo, Luis; Kalb, Samuel; Reyes, Phillip M; Chang, Steve W; Crawford, Neil R

2013-04-15

Seven different combinations of posterior screw fixation, with or without interbody support, were compared in vitro using nondestructive flexibility tests. To study the biomechanical behavior of a new cortical screw (CS) fixation construct relative to the traditional pedicle screw (PS) construct. The CS is an alternative to the PS for posterior fixation of the lumbar spine. The CS trajectory is more sagittally and cranially oriented than the PS, being anchored in the pars interarticularis. Like PS fixation, CS fixation uses interconnecting rods fastened with top-locking connectors. Stability after bilateral CS fixation was compared with stability after bilateral PS fixation in the setting of intact disc and with direct lateral interbody fixation (DLIF) or transforaminal lateral interbody fixation (TLIF) support. Standard nondestructive flexibility tests were performed in cadaveric lumbar specimens, allowing non-paired comparisons of specific conditions from 28 specimens (4 groups of 7) within a larger experiment of multiple hardware configurations. Condition tested and group from which results originated were as follows: (1) intact (all groups); (2) with L3-L4 bilateral PS-rods (group 1); (3) with bilateral CS-rods (group 2); (4) with DLIF (group 3); (5) with DLIF + CS-rods (group 4); (6) with DLIF + PS-rods (group 3); (7) with TLIF + CS-rods (group 2), and (8) with TLIF + PS-rods (group 2). To assess spinal stability, the mean range of motion, lax zone, and stiff zone at L3-L4 were compared during flexion-extension, lateral bending, and axial rotation. With intact disc, stability was equivalent after PS-rod and CS-rod fixation, except that PS-rod fixation was stiffer during axial rotation. With DLIF support, there was no significant difference in stability between PS-rod and CS-rod fixation. With TLIF support, PS-rod fixation was stiffer than CS-rod fixation during lateral bending. Bilateral CS-rod fixation provided about the same stability in cadaveric specimens

NSRR experiment with un-irradiated uranium-zirconium hydride fuel. Design, fabrication process and inspection data of test fuel rod

Energy Technology Data Exchange (ETDEWEB)

Sasajima, Hideo; Fuketa, Toyoshi; Ishijima, Kiyomi; Kuroha, Hiroshi; Ikeda, Yoshikazu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Aizawa, Keiichi

1998-08-01

An experiment plan is progressing in the Nuclear Safety Research Reactor (NSRR) to perform pulse-irradiation with uranium-zirconium hydride (U-ZrH{sub x}) fuel. This fuel is widely used in the training research and isotope production reactor of GA (TRIGA). The objectives of the experiment are to determine the fuel rod failure threshold and to investigate fuel behavior under simulated reactivity initiated accident (RIA) conditions. This report summarizes design, fabrication process and inspection data of the test fuel rods before pulse-irradiation. The experiment with U-ZrH{sub x} fuel will realize precise safety evaluation, and improve the TRIGA reactor performance. The data to be obtained in this program will also contribute development of next-generation TRIGA reactor and its safety evaluation. (author)

PHEBUS/test-218, Behaviour of a Fuel Rod Bundle during a Large Break LOCA Transient with a two Peaks Temperature History

International Nuclear Information System (INIS)

1987-01-01

1 - Description of test facility: PHEBUS test facility operated at CEA Research Center Cadarache consists of a pressurized circuit involving pumps, heat exchangers and a blowdown tank - 25 nuclear fuel rod bundle, coupled to a separate driver core; - active length 0.8 m, cosine axial power profile; - pressurized and un-pressurized fuel rods; - controlled cooling conditions at the bundle inlet (blowdown, refill and reflood period); - de-pressurized test rig volume 0.22 m 3 . The following 'as measured' boundary conditions (B.C.) were offered to participants as options with decreasing challenge to their analytical approach: Boundary conditions B.C.0: - full thermal-hydraulic analysis of PHEBUS test rig (was not recommended). Boundary conditions B.C.1: - thermal power level of fuel bundle; - fluid inlet conditions to bundle section. Boundary conditions B.C.2: - local cladding temperatures of rods; - heat transfer coefficients. Boundary conditions B.C.3: - cladding temperatures of rods; - internal pressure of rods. 2 - Description of test: Post-test investigation into the response of a nuclear fuel bundle to a large break loss of coolant accident with respect to - local fuel temperatures, - cladding strain at the time of burst, - time to burst and under given thermal-hydraulic boundary conditions of PHEBUS-test 218

MABEL-2: a code to analyse cladding deformation in a loss-of-coolant accident

International Nuclear Information System (INIS)

Bowring, R.W.; Cooper, C.A.

1982-04-01

MABEL-2 has been developed to predict the extent of cladding deformation in PWR fuel rods during a loss of coolant accident. The user notes describe how to run MABEL. They include case preparation and input data, the job control language, a description of the output tables available, and aids to debugging. The input data and results for two sample cases are given. (U.K.)

Review of accident analyses of RB experimental reactor

International Nuclear Information System (INIS)

Pesic, M.

2003-01-01

The RB reactor is a uranium fuel heavy water moderated critical assembly that has been put and kept in operation by the VINCA Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since April 1958. The first complete Safety Analysis Report of the RB reactor was prepared in 1961/62; yet, the first accident analysis had been made in late 1958 with the aim to examine a power transition and the total equivalent doses received by the staff during the reactivity accident that occurred on October 15, 1958. Since 1960, the RB reactor has been modified a few times. Beside the initial natural uranium metal fuel rods, new types of fuel (TVR-S types of Russian origin) consisting of 2% enriched uranium metal and 80% enriched U0 2 , dispersed in aluminum matrix, have been available since 1962 and 1976, respectively. Modifications of the control and safety systems of the reactor were made occasionally. Special reactor cores were designed and constructed using all three types of fuel elements, as well as the coupled fast-thermal ones. The Nuclear Safety Committee of the VINCA Institute, an independent regulator)' body, approved for usage all these modifications of the RB reactor on the basis of the Preliminary Safety' Analysis Reports, which, beside proposed technical modifications and new regulation rules, included safety analyses of various possible accidents. A special attention was given (and a new safety methodology was proposed) to thorough analyses of the design-based accidents related to the coupled fast-thermal cores that included central zones of the reactor filled by the fuel elements without any moderator. In this paper, an overview of some accidents, methodologies and computation tools used for the accident analyses of the RB reactor is given. (author)

Review of accident analyses of RB experimental reactor

Directory of Open Access Journals (Sweden)

Pešić Milan P.

2003-01-01

Full Text Available The RB reactor is a uranium fuel heavy water moderated critical assembly that has been put and kept in operation by the VTNCA Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since April 1958. The first complete Safety Analysis Report of the RB reactor was prepared in 1961/62 yet, the first accident analysis had been made in late 1958 with the aim to examine a power transition and the total equivalent doses received by the staff during the reactivity accident that occurred on October 15, 1958. Since 1960, the RB reactor has been modified a few times. Beside the initial natural uranium metal fuel rods, new types of fuel (TVR-S types of Russian origin consisting of 2% enriched uranium metal and 80% enriched UO2 dispersed in aluminum matrix, have been available since 1962 and 1976 respectively. Modifications of the control and safety systems of the reactor were made occasionally. Special reactor cores were designed and constructed using all three types of fuel elements as well as the coupled fast-thermal ones. The Nuclear Safety Committee of the VINĆA Institute, an independent regulatory body, approved for usage all these modifications of the RB reactor on the basis of the Preliminary Safety Analysis Reports, which, beside proposed technical modifications and new regulation rules, included safety analyses of various possible accidents. A special attention was given (and a new safety methodology was proposed to thorough analyses of the design-based accidents related to the coupled fast-thermal cores that included central zones of the reactor filled by the fuel elements without any moderator. In this paper, an overview of some accidents, methodologies and computation tools used for the accident analyses of the RB reactor is given.

Freely suspended rod fall dampener, especially for control rod of liquid-cooled nuclear reactor

International Nuclear Information System (INIS)

Becvar, J.; Saroch, V.

1977-01-01

A shock absorber is described whose advantage is that the space required for the movement of the shock absorber in the operating travel of the system suspension rod-control rod bundle may be reduced. The design allows the automatic disconnection of the system and the removal of the suspension rod from the reactor without dismantling. The braking force reaction is transmitted to the structure above the core. The system fall energy is absorbed on the side of the suspension rod which has a bigger mass. (J.B.)

Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

Energy Technology Data Exchange (ETDEWEB)

Baek J.; Diamond D.; Cuadra, A.; Hanson, A.L.; Cheng, L-Y.; Brown, N.R.

2012-09-30

Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a model of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.

RodPilot{sup R} - The Innovative and Cost-Effective Digital Control Rod Drive Control System for PWRs

Energy Technology Data Exchange (ETDEWEB)

Baron, Clemens [AREVA NP GmbH, NLEE-G, Postfach 1199, 91001 Erlangen (Germany)

2008-07-01

With RodPilot, AREVA NP offers an innovative and cost-effective system for controlling control rods in Pressurized Water Reactors. RodPilot controls the three operating coils of the control rod drive mechanism (lift, moveable gripper and stationary gripper coil). The rods are inserted into or withdrawn from the core as required by the Reactor Control System. The system combines modern components, state-of-the-art logic and a proven electronic control rod drive control principle to provide enhanced reliability and lower maintenance costs. (author)

Analysis of hypothetical loss-of-control-arm accidents in HIFAR

International Nuclear Information System (INIS)

Connolly, J.W.; Clark, N.

1986-11-01

The reactor power transient produced in the HIFAR materials testing reactor upon severance of a central coarse control arm connecting rod and the subsequent pivoting of the arm out of the core has been calculated for a range of reactor conditions likely to be encountered in normal operation. It is concluded that as long as the remaining arms of the control arm bank can be relied on to suppress the post power peak oscillations in power, the reactor will withstand the consequences of such an accident

Study of heat and mass transfer phenomena in fuel assembly models under accident conditions

International Nuclear Information System (INIS)

Yefanov, A.D.; Kalyakin, C.G.; Loshchinin, V.M.; Pomet'ko, R.S.; Sergeev, V.V.; Shumsky, R.V.

1996-01-01

The majority of the material in support of the thermal - hydraulic safety of WWER core was obtained on single - assembly models containing a relatively small number of elements - heater rods. Upgrading the requirements to the reactor safety leads to the necessity for studying phenomena in channels representing the cross - sectional core dimensions and non - uniform radial power generation. Under such conditions, the contribution of natural convection can be significant in some core zones, including the occurrence of reverse flows and interchannel instability. These phenomena can have an important influence on heat transfer processes. Such influence is especially drastical under accident conditions associated with ceasing the forced circulation over the circuit. A number of urgent reactor safety problems at low operating parameters is related with the computer code verification and certification. One of the important trends in the reactor safety research is concerned with the rod bundle reflooding and verificational calculations of this phenomenon. To assess the water cooled reactor safety, the best fit computer codes are employed, which make it possible to simulate accident and transient operating conditions in a reactor installation. One of the most widely known computer codes is the RELAP5/MOD3 Code. The paper presents the comparison of the results calculated using this computer code with the test data on 4 - rod bundle quenching, which were obtained at the SSCRF-IPPE. Recently, the investigations on the steam - zirconium reaction kinetics have been performed at the SSCFR-IPPE and are being presently performed for the purpose of developing new and verifying available computer codes. (author). 3 refs, 6 figs

Control-rod driving mechanism

International Nuclear Information System (INIS)

Jodoi, Takashi.

1976-01-01

Purpose: To prevent falling of control rods due to malfunction. Constitution: The device of the present invention has a scram function in particular, and uses principally a fluid pressure as a scram accelerating means. The control rod is held by upper and lower holding devices, which are connected by a connecting mechanism. This connecting mechanism is designed to be detachable only at the lower limit of driving stroke of the control rod so that there occurs no erroneous scram resulting from careless disconnection of the connecting mechanism. Further, scramming operation due to own weight of the scram operating portion such as control rod driving shaft may be effected to increase freedom. (Kamimura, M.)

REACTOR CONTROL ROD OPERATING SYSTEM

Science.gov (United States)

Miller, G.

1961-12-12

A nuclear reactor control rod mechanism is designed which mechanically moves the control rods into and out of the core under normal conditions but rapidly forces the control rods into the core by catapultic action in the event of an emergency. (AEC)

Proceedings of the Second Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs, 23-25 September 2014, OECD-NEA HQ

International Nuclear Information System (INIS)

Massara, S.; ); Bragg-Sitton, Shannon; Pasamehmetoglu, K.; Yang, Jae Ho; Dolley, Evan J.; Rebak, Raul B.; Sowder, Andrew; Cheng, Bo; Kurata, Masaki; Van Nieuwenhove, Rudi; Li, R.; McClellan, Ken; Nelson, Andy; Carmack, Jon; Harp, Jason; Finck, Phillip; ); Kakicuhi, K.

2014-09-01

Under the guidance of the OECD-NEA Nuclear Science Committee, the expert group acts as a forum for scientific and technical information exchange on advanced light water reactor (LWR) fuels with enhanced accident tolerance. The expert group focusses on the fundamental properties and behaviour under normal operations and accident conditions for advanced core materials and components (fuels, cladding, control rods, etc.). The materials considered are applicable to Gen II and Gen III Light Water Reactors, as well as Gen III+ reactors under construction. The objective of the expert group is to define and coordinate a programme of work to help advance the scientific knowledge needed to provide the technical underpinning for the development of advanced LWR fuels with enhanced accident tolerance compared to currently used zircaloy/UO 2 fuel systems, as well as other non-fuel core components with important roles in LWR performance under accident conditions. This document brings together the available presentations (slides) given at the Second Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs. Content: 1 - Proposed Agenda; 2 - Expert Group meeting - 23 September 2014: - Introduction and background (S. Massara, OECD-NEA) - Expected outcomes from the TFs meetings scheduled on 24-25 September (K. Pasamehmetoglu, EG Chair, INL); 3 - Task Force 1 (Systems assessment) meeting - 24 September 2014: - Metrics for the Evaluation of LWR Accident Tolerant Fuel (S. Bragg-Sitton, INL); 4 - Task Force 2 (Cladding/core materials) meeting - 24 September 2014: - Summary on SiC Task Force 2 (Clad) meeting (J.H. Yang, KAERI); - Accident Tolerant Advanced Steels Cladding for Commercial Light Water Reactors (E. Dolley, GE); - Molybdenum-Alloy Fuel Cladding Development and Testing - Update from April 2014 NEA ATF Meeting (A. Sowder, EPRI); - Accident Tolerant Control Rod Development in Japan (M. Kurata, JAEA); - IFA-774: The first in-pile test with coated fuel rods (R. Van

The analysis of RWAP(Rod Withdrawal at Power) using the KEPRI methodology

International Nuclear Information System (INIS)

Yang, C. K.; Kim, Y. H.

2001-01-01

KEPRI developed new methodology which was based on RASP(Reactor Analysis Support Package). In this paper, The analysis of RWAP(Rod Withdrawal at Power) accident which can result in reactivity and power distribution anomaly was performed using the KEPRI methodology. The calculation describes RWAP transient and documents the analysis, including the computer code modeling assumptions and input parameters used in the analysis. To validity for the new methodology, the result of calculation was compared with FSAR. As compared with FSAR, result of the calculation using the KEPRI Methodology is similar to FSAR's. And result of the sensitivity of postulated parameters were similar to the existing methodology

Control rod velocity limiter

International Nuclear Information System (INIS)

Cearley, J.E.; Carruth, J.C.; Dixon, R.C.; Spencer, S.S.; Zuloaga, J.A. Jr.

1986-01-01

This patent describes a velocity control arrangement for a reciprocable, vertically oriented control rod for use in a nuclear reactor in a fluid medium, the control rod including a drive hub secured to and extending from one end therefrom. The control device comprises: a toroidally shaped control member spaced from and coaxially positioned around the hub and secured thereto by a plurality of spaced radial webs thereby providing an annular passage for fluid intermediate the hub and the toroidal member spaced therefrom in coaxial position. The side of the control member toward the control rod has a smooth generally conical surface. The side of the control member away from the control rod is formed with a concave surface constituting a single annular groove. The device also comprises inner and outer annular vanes radially spaced from one another and spaced from the side of the control member away from the control rod and positioned coaxially around and spaced from the hub and secured thereto by spaced radial webs thereby providing an annular passage for fluid intermediate the hub and the vanes. The vanes are angled toward the control member, the outer edge of the inner vane being closer to the control member and the inner edge of the outer vane being closer to the control member. When the control rod moves in the fluid in the direction toward the drive hub the vanes direct a flow of fluid turbulence which provides greater resistance to movement of the control rod in the direction toward the drive hub than in the other direction

MABEL-2D: a code to analyse cladding deformation in a loss-of-coolant accident. Part 2

International Nuclear Information System (INIS)

Bowring, R.W.

1985-08-01

The MABEL series of codes is being developed at Harwell to predict the extent of cladding deformation (ballooning) in pressurized water reactor fuel rods during a loss of coolant accident. MABEL - 2D is an updated version of MABEL - 2C. These are user notes for MABEL - 2D (which is described in a separate report AEEW - R1979). They describe the input data specification; the use of the restart facility; debug printing and quick-running sample problems. The input data are divided into rod data, thermal hydraulic data and creep data. There is an input data flow chart. The main appendix gives the detailed input data specification. (U.K.)

Test Facility Construction for Flow Visualization on Mixing Flow inside Subchannels of PWR Rod Bundle

Energy Technology Data Exchange (ETDEWEB)

Kim, Seok; Jeon, Byong-Guk; Youn, Young-Jung; Choi, Hae-Seob; Euh, Dong-Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Flow inside rod bundles has a similarity with flow in porous media. To ensure thermal performance of a nuclear reactor, detailed information of the heat transfer and turbulent mixing flow phenomena taking place within the subchannels is required. The subchannel analysis is one of the key thermal-hydraulic calculations in the safety analysis of the nuclear reactor core. At present, subchannel computer codes are employed to simulate fuel elements of nuclear reactor cores and predict the performance of cores under normal operating and hypothetical accident conditions. The ability of these subchannels codes to predict both the flow and enthalpy distribution in fuel assemblies is very important in the design of nuclear reactors. Recently, according to the modern tend of the safety analysis for the nuclear reactor, a new component scale analysis code, named CUPID, and has been developed in KAERI. The CUPID code is based on a two-fluid and three-field model, and both the open and porous media approaches are incorporated. The PRIUS experiment has addressed many key topics related to flow behaviour in a rod bundle. These issues are related to the flow conditions inside a nuclear fuel element during normal operation of the plant or in accident scenarios. From the second half of 2016, flow visualization will be performed by using a high speed camera and image analysis technique, from which detailed information for the two-dimensional movement of single phase flow is quantified.

Control rod position control device

International Nuclear Information System (INIS)

Ubukata, Shinji.

1997-01-01

The present invention provides a control rod position control device which stores data such as of position signals and driving control rod instruction before and after occurrence of abnormality in control for the control rod position for controlling reactor power and utilized the data effectively for investigating the cause of abnormality. Namely, a plurality of individual control devices have an operation mismatching detection circuit for outputting signals when difference is caused between a driving instruction given to the control rod position control device and the control rod driving means and signals from a detection means for detecting an actual moving amount. A general control device collectively controls the individual control devices. In addition, there is also disposed a position storing circuit for storing position signals at least before and after the occurrence of the control rod operation mismatching. With such procedures, the cause of the abnormality can be determined based on the position signals before and after the occurrence of control rod mismatching operation stored in the position storing circuit. Accordingly, the abnormality cause can be determined to conduct restoration in an early stage. (I.S.)

Control rod selecting and driving device

International Nuclear Information System (INIS)

Isobe, Hideo.

1981-01-01

Purpose: To simultaneously drive a predetermined number of control rods in a predetermined mode by the control of addresses for predetermined number of control rods and read or write of driving codified data to and from the memory by way of a memory controller. Constitution: The system comprises a control rod information selection device for selecting predetermined control rods from a plurality of control rods disposed in a reactor and outputting information for driving them in a predetermined mode, a control rod information output device for codifying the information outputted from the above device and outputting the addresses to the predetermined control rods and driving mode coded data, and a driving device for driving said predetermined control rods in a predetermined mode in accordance with the codified data outputted from the above device, said control rod infromation output device comprising a memory device capable of storing a predetermined number of the codified data and a memory control device for storing the predetermined number of data into the above memory device at a predetermined timing while successively outputting the thus stored predetermined number of data at a predetermined timing. (Seki, T.)

NEUTRONIC REACTOR CONTROL ROD DRIVE APPARATUS

Science.gov (United States)

Oakes, L.C.; Walker, C.S.

1959-12-15

ABS>A suspension mechanism between a vertically movable nuclear reactor control rod and a rod extension, which also provides information for the operator or an automatic control signal, is described. A spring connects the rod extension to a drive shift. The extension of the spring indicates whether (1) the rod is at rest on the reactor, (2) the rod and extension are suspended, or (3) the extension alone is suspended, the spring controlling a 3-position electrical switch.

Ultrasonographic ejection fraction of normal gallbladder

Energy Technology Data Exchange (ETDEWEB)

Park, Jin Hun; Kim, Seung Yup; Park, Yaung Hee; Kang, Ik Won; Yoon, Jong Sup [Hangang Sacred Heart Hospital, Halym College, Chuncheon (Korea, Republic of)

1984-06-15

Real-time ultrasonography is a simple, accurate, noninvasive and potentially valuable means of studying gallbladder size and emptying. The authors calculated ultrasonographically the ejection fraction of 80 cases of normally functioning gallbladder on oral cholecystography, from June 1983 to April 1984, at the department of radiology, Hangang Sacred Heart Hospital. The results were obtained as follows; 1. Ultrasonographic Ejection Fraction at 30 minutes after the fatty meal was 73.1{+-}16.85. 2. There was no significant difference in age and sex, statistically.

BISON Fuel Performance Analysis of IFA-796 Rod 3 & 4 and Investigation of the Impact of Fuel Creep

Energy Technology Data Exchange (ETDEWEB)

Wirth, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sweet, Ryan T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-08-01

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace the currently used zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromiumaluminum (FeCrAl) alloys because they exhibit slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and slow cladding consumption in the presence of high temperature steam. These alloys should also exhibit increased “coping time” in the event of an accident scenario by improving the mechanical performance at high temperatures, allowing greater flexibility to achieve core cooling. As a continuation of the development of these alloys, in-reactor irradiation testing of FeCrAl cladded fuel rods has started. In order to provide insight on the possible behavior of these fuel rods as they undergo irradiation in the Halden Boiling Water Reactor, engineering analysis has been performed using FeCrAl material models implemented into the BISON fuel performance code. This milestone report provides an update on the ongoing development of modeling capability to predict FeCrAl cladding fuel performance and to provide an early look at the possible behavior of planned in-reactor FeCrAl cladding experiments. In particular, this report consists of two separate analyses. The first analysis consists of fuel performance simulations of IFA-796 rod 4 and two segments of rod 3. These simulations utilize previously implemented material models for the C35M FeCrAl alloy and UO2 to provide a bounding behavior analysis corresponding to variation of the initial fuel cladding gap thickness within the fuel rod. The second analysis is an assessment of the fuel and cladding stress states after modification of the fuel creep model that is currently implemented in the BISON fuel performance code. Effects from modifying the fuel creep model were identified for the BISON simulations

RODDRP - A FORTRAN program for use in control rod calibration by the rod drop method

International Nuclear Information System (INIS)

Wilson, W.E.

1972-01-01

The different methods to measure reactivity which are applicable to control rod calibration are discussed. They include: 1) the positive period method, 2) the rod drop method, 3) the source-jerk method, 4) the rod oscillation method, and 5) the pulsed neutron method. The instrument setup used at WSU for rod drop measurements is presented. To speed up the analysis of power fall-off trace, a FORTRAN IV program called RODDRP was written to simultaneously solve the in-hour equation and relative neutron flux. The procedure for calculating the worth of the rod that produced the power trace is given. The reactivity for each time relative flux point is obtained. Conclusions about the status of the equipment are made

Parametric study of recriticality in a boiling water reactor severe accident

International Nuclear Information System (INIS)

Shamoun, B.I.; Witt, R.J.

1994-01-01

Recriticality is possible in a severe accident if unborated or low boron concentration water is added to a damaged core after control rod melting but before fuel melting. Recriticality in a severe accident in a boiling water reactor was parametrically investigated using the TWODANT code. Eigenvalue calculations for a unit central fuel cell with reflective boundary conditions were performed by solving the two-dimensional multigroup steady-state Boltzman transport equation using TWODANT. Two sets of calculations were performed in this work. The first set of calculations was carried out under three types of normal operating conditions to provide reference values for the accident calculations: (a) cold rodded condition, (b) cold unrodded condition, and (c) hot full-power condition. The eigenvalues at these conditions were found to be 1.055, 1.208, and 1.098, respectively. The second set of calculations was carried out after the melting of the control element and during the reflood phase, under the following reflood conditions: (a) reflood with unborated water and (b) reflood with borated water. For the reflood case with unborated water, five values of void fractions were considered (100, 60, 40, 20, and 0%). Decreasing void fractions represent greater refill levels during the reflood process. The system pressure was taken to be 7 MPa, while the moderator temperature was set to 560 K. Plotting the eigenvalue compared with the fraction of control materials lost indicates recriticality is only possible if nearly 100% of the control material is lost from the core. Eigenvalue calculations were repeated for short- and long-term recovery conditions of the reflood phase corresponding to maximum moderator density at 4 MPa pressure and 525 K moderator temperature and for 1 MPa pressure and 325 K moderator temperature, respectively. Recriticality was again observed to be a concern only after losing 95% ore more of control materials from the unit cell

Thermalhydraulic behavior of electrically heated rods during critical heat flux transients

International Nuclear Information System (INIS)

Lima, Rita de Cassia Fernandes de

1997-01-01

In nuclear reactors, the occurrence of critical heat flux leads to fuel rod overheating with clad fusion and radioactive products leakage. To predict the effects of such phenomenon, experiments are performed utilizing heated rods to simulate operational and accidental conditions of nuclear fuel rods, with special attention to the phenomenon of boiling crisis. The use of mechanisms which detect the abrupt temperature rise allows the electric power switch off. These facts prevent the test section from damage. During the critical heat flux phenomenon the axial heat conduction becomes very important. The study of the dryout and rewetting fronts yields the analysis, planning and following of critical heat flux experiments. These facts are important during the reflooding of nuclear cores at severe accidents. In the present work it is performed a theoretical analysis of the drying and rewetting front propagation during a critical heat flux experiment, starting with the application of an electrical power step or power slope from steady state condition. After the occurrence of critical heat flux, it is predicted the drying front propagation. After a few seconds, a power cut is considered and the rewetting front behavior is analytically observed. In all these transients the coolant pressure is 13,5 MPa. For one of them, comparisons are done with a pressure of 8,00 MPa. Mass flow and enthalpy influences on the fronts velocities are also analysed. These results show that mass flow has more importance on the drying front velocities whereas the pressure alters strongly the rewetting ones. (author)

External attachment of titanium sheathed thermocouples to zirconium nuclear fuel rods for the LOFT reactor

International Nuclear Information System (INIS)

Welty, R.K.

1980-01-01

The Exxon Nuclear Company, Inc., acting as a Subcontractor to EG and G Idaho Inc., Idaho National Engineering Laboratory, Idaho Falls, Idaho, has developed a welding process to attach titanium sheathed thermocouples to the outside of the zircaloy clad fuel rods. The fuel rods and thermocouples are used to test simulated loss-of-coolant accident (LOCA) conditions in a pressurized water reactor (LOFT Reactor, Idaho National Laboratory). A laser beam was selected as the optimum welding process because of the extremely high energy input per unit volume that can be achieved allowing local fusion of a small area irrespective of the difference in material thickness to be joined. A commercial pulsed laser and energy control system was installed along with specialized welding fixtures. Laser room facility requirements and tolerances were established. Performance qualifications, and detailed welding procedures were also developed. Product performance tests were conducted to assure that engineering design requirements could be met on a production basis

Simulation of vibration modes of the fuel rod damaged due to the grid-to-rod fretting wear

International Nuclear Information System (INIS)

Kim, Kyu Tae; Kim, Kyeong Koo; Jang, Young Ki; Lee, Kyou Seok

1997-01-01

The flow-induced fuel fretting wear observed in some PWRs mainly proceeds in the grid-to-rod contact positions. The grid-to-rod fretting wear in the PWR fuel assembly depends on grid-to-rod gap size, its axial profile and flow-induced vibration. This paper describes the GRIDFORCE program which generates the axially dependent grid-to-rod gap size as a function of burnup. The axially dependent grid-to-rod gap profiles are employed to predict the fuel rod vibration mode shapes by the ANSYS code. With the help of the Paidousis empirical formula, this paper also calculates the fuel rod vibration amplitudes under various supporting conditions, which indicates that the increase of the number of unsupported mid-grids will increase the fuel rod vibration amplitude. On the other hand, the comparison of the predicted vibration mode shapes and the observed mid-grid fretting wear pattern indicates that the 1st and 6th vibration mode shapes under the supporting inactive condition at the mid-grids can simulate the observed mid-grid fretting wear profile. This paper also proposes design guidelines against the grid-to-rod fretting wear. (author). 3 refs., 8 figs

Reactivity initiated accident analyses for the safety assessment of upgraded JRR-3

International Nuclear Information System (INIS)

Harami, Taikan; Uemura, Mutsumi; Ohnishi, Nobuaki

1984-08-01

JRR-3, currently a heavy water moderated and cooled 10 MW reactor, is to be upgraded to a light water moderated and cooled, heavy water reflected 20 MW reactor. This report describes the analytical results of reactivity initiated accidents for the safety assessment of upgraded JRR-3. The following five cases have been selected for the assessment; (1) uncontrolled control rod withdrawal from zero power, (2) uncontrolled control rod withdrawal from full power, (3) removal of irradiation samples, (4) increase of primary coolant flow, (5) failure of heavy water tank. Parameter studies have been made for each of the above cases to cover possible uncertainties. All analyses have been made by a computer code EUREKA-2. The results show that the safety criteria for upgraded JRR-3 are all met and the adequacy of the design is confirmed. (author)

Automated nuclear fuel rod pattern loading system

International Nuclear Information System (INIS)

Lambert, D.V.; Nylund, T.W.; Byers, J.W.; Haley, D.E. Jr.; Cioffi, J.V.

1991-01-01

This patent describes a method for loading fuel rods in a desired pattern. It comprises providing a supply of fuel rods of known enrichments; providing a magazine defining a matrix of elongated slots open at their forward ends for receiving fuel rods; defining a fuel rod feed path; receiving successively one at a time along the feed path fuel rods selected from the supply thereof; verifying successively one at a time along the feed path the identity of the selected fuel rods, the verifying including blocking passage of each selected fuel rod along the feed path until the identity of each selected fuel rod is confirmed as correct; feeding to the magazine successively one at a time along the feed path the selective and verified fuel rods; and supporting and moving the magazine along X-Y axes to successively align one at a time selected ones of the slots with the feed path for loading in the magazine the successive fuel rods in a desired enrichment pattern

Control rod drives

International Nuclear Information System (INIS)

Ikakura, Hiroaki.

1986-01-01

Purpose: To enable to direct disconnection of control rods upon abnormal temperature rise in the reactor thereby improve the reliability for the disconnecting operation in control rod drives for FBR type reactors upon emergency. Constitution: A diaphragm is disposed to the upper opening of a sealing vessel inserted to the hollow portion of an electromagnet and a rod is secured to the central position of the upper surface. A spring contacts are attached by way of an insulator to the inner surface at the lower portion of an extension pipe and connected with cables for supplying electric power sources respectively to a magnet. If the temperature in the reactor abnormally rises, liquid metals in the sealing vessel are expanded tending to extend the bellows downwardly. However, since they are attracted by the electromagnet, the thermal expansion of the liquid metals exert on the diaphragm prior to the bellows. Thus, the switch between the spring contacts is made open to attain the deenergized state to thereby disconnect the control rod and shutdown the neclear reactor. (Horiuchi, T.)

Method for compacting spent nuclear reactor fuel rods

International Nuclear Information System (INIS)

Wachter, W.J.

1988-01-01

In a nuclear reactor system which requires periodic physical manipulation of spent fuel rods, the method of compacting fuel rods from a fuel rod assembly is described. The method consists of: (1) removing the top end from the fuel rod assembly; (2) passing each of multiple fuel rod pulling elements in sequence through a fuel rod container and thence through respective consolidating passages in a fuel rod directing chamber; (3) engaging one of the pulling elements to the top end of each of the fuel rods; (4) drawing each of the pulling elements axially to draw the respective engaged fuel rods in one axial direction through the respective the passages in the chamber to thereby consolidate the fuel rods into a compacted configuration of a cross-sectional area smaller than the cross-sectional area occupied thereby within the fuel rod assembly; and (5) drawing all of the engaged fuel rods concurrently and substantially parallel to one another in the one axial direction into the fuel rod container while maintaining the compacted configuration whereby the fuel rods are aligned within the container in a fuel rod density of the the fuel rod assembly

Do centrioles generate a polar ejection force?

Science.gov (United States)

Wells, Jonathan

2005-01-01

A microtubule-dependent polar ejection force that pushes chromosomes away from spindle poles during prometaphase is observed in animal cells but not in the cells of higher plants. Elongating microtubules and kinesin-like motor molecules have been proposed as possible causes, but neither accounts for all the data. In the hypothesis proposed here a polar ejection force is generated by centrioles, which are found in animals but not in higher plants. Centrioles consist of nine microtubule triplets arranged like the blades of a tiny turbine. Instead of viewing centrioles through the spectacles of molecular reductionism and neo-Darwinism, this hypothesis assumes that they are holistically designed to be turbines. Orthogonally oriented centriolar turbines could generate oscillations in spindle microtubules that resemble the motion produced by a laboratory vortexer. The result would be a microtubule-mediated ejection force tending to move chromosomes away from the spindle axis and the poles. A rise in intracellular calcium at the onset of anaphase could regulate the polar ejection force by shutting down the centriolar turbines, but defective regulation could result in an excessive force that contributes to the chromosomal instability characteristic of most cancer cells.

Control rod drives

International Nuclear Information System (INIS)

Asano, Hiromitsu.

1979-01-01

Purpose: To drive control rods at an optimum safety speed corresponding to the reactor core output. Constitution: The reactor power is detected by a neutron detector and the output signal is applied to a process computer. The process computer issues a signal representing the reactor core output, which is converted through a function generator into a signal representing the safety speed of control rods. The converted signal is further supplied to a V/F converter and converted into a pulse signal. The pulse signal is inputted to a step motor driving circuit, which actuates a step motor to operate the control rods always at a safety speed corresponding to the reactor core power. (Furukawa, Y.)

Magnetohydrodynamic simulations of the ejection of a magnetic flux rope

Science.gov (United States)

Pagano, P.; Mackay, D. H.; Poedts, S.

2013-06-01

Context. Coronal mass ejections (CME's) are one of the most violent phenomena found on the Sun. One model to explain their occurrence is the flux rope ejection model. In this model, magnetic flux ropes form slowly over time periods of days to weeks. They then lose equilibrium and are ejected from the solar corona over a few hours. The contrasting time scales of formation and ejection pose a serious problem for numerical simulations. Aims: We simulate the whole life span of a flux rope from slow formation to rapid ejection and investigate whether magnetic flux ropes formed from a continuous magnetic field distribution, during a quasi-static evolution, can erupt to produce a CME. Methods: To model the full life span of magnetic flux ropes we couple two models. The global non-linear force-free field (GNLFFF) evolution model is used to follow the quasi-static formation of a flux rope. The MHD code ARMVAC is used to simulate the production of a CME through the loss of equilibrium and ejection of this flux rope. Results: We show that the two distinct models may be successfully coupled and that the flux rope is ejected out of our simulation box, where the outer boundary is placed at 2.5 R⊙. The plasma expelled during the flux rope ejection travels outward at a speed of 100 km s-1, which is consistent with the observed speed of CMEs in the low corona. Conclusions: Our work shows that flux ropes formed in the GNLFFF can lead to the ejection of a mass loaded magnetic flux rope in full MHD simulations. Coupling the two distinct models opens up a new avenue of research to investigate phenomena where different phases of their evolution occur on drastically different time scales. Movies are available in electronic form at http://www.aanda.org

Individual nuclear fuel rod weighing system

International Nuclear Information System (INIS)

Fogg, J. L.; Howell, C. A.; Smith, J. H.; Vining, G. E.

1985-01-01

An individual nuclear fuel rod weighing system for rods carried on a tray which moves along a materials handling conveyor. At a first tray position on the conveyor, a lifting device raises the rods off the tray and places them on an overhead ramp. A loading mechanism conveys the rods singly from the overhead ramp onto an overhead scale for individual weighing. When the tray is at a second position on the conveyor, a transfer apparatus transports each weighed rod from the scale back onto the tray

Measuring device for control rod driving time

International Nuclear Information System (INIS)

Tanaka, Kazuhiko; Hanabusa, Masatoshi.

1993-01-01

The present invention concerns a measuring device for control driving time having a function capable of measuring a selected control rod driving time and measuring an entire control rod driving time simultaneously. A calculation means and a store means for the selected rod control rod driving time, and a calculation means and a store means for the entire control rod driving time are disposed individually. Each of them measures the driving time and stores the data independent of each other based on a selected control rod insert ion signal and an entire control rod insertion signal. Even if insertion of selected and entire control rods overlaps, each of the control rod driving times can be measured reliably to provide an advantageous effect capable of more accurately conducting safety evaluation for the nuclear reactor based on the result of the measurement. (N.H.)

Growth and Morphology of Rod Eutectics

Energy Technology Data Exchange (ETDEWEB)

Jing Teng; Shan Liu; R. Trivedi

2008-03-17

The formation of rod eutectic microstructure is investigated systematically in a succinonitrile-camphor alloy of eutectic composition by using the directional solidification technique. A new rod eutectic configuration is observed in which the rods form with elliptical cylindrical shape. Two different orientations of the ellipse are observed that differ by a 90{sup o} rotation such that the major and the minor axes are interchanged. Critical experiments in thin samples, where a single layer of rods forms, show that the spacing and orientation of the elliptic rods are governed by the growth rate and the sample thickness. In thicker samples, multi layers of rods form with circular cross-section and the scaling law between the spacing and velocity predicted by the Jackson and Hunt model is validated. A theoretical model is developed for a two-dimensional array of elliptical rods that are arranged in a hexagonal or a square array, and the results are shown to be consistent with the experimental observations. The model of elliptic rods is also shown to reduce to that for the circular rod eutectic when the lengths of the two axes are equal, and to the lamellar eutectic model when one of the axes is much larger than the other one.

Status of rod consolidation, 1988

International Nuclear Information System (INIS)

Bailey, W.J.

1989-01-01

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

Inspecting method for fuel rods

International Nuclear Information System (INIS)

Watanabe, Masaaki; Kogure, Sumio.

1976-01-01

Purpose: To precisely detect the response of flaw in clad tube and submerged fuel pellets from a relationship between the surface of fuel rod and internal signal. Constitution: Ultrasonic reflected waves from the surface of fuel rods and the interior are detected and either one of fuel rod or ultrasonic flaw detecting contact is rotated to thereby precisely detect the response of the flaw of clad tube and submerged fuel pellets from a relationship between said surface and the interior. It will be noted that the ultrasonic flaw detecting contact used is of the line-focus type, the incident angle of ultrasonic wave from the ultrasonic flaw detecting contact relative to the fuel rod is the angle of skew, that is, the ultrasonic flaw detecting contact is not perpendicular to a center axis of the fuel rod but is slightly displace. That is, the use of the aforesaid contact may facilitate discrimination between the surface flaw of the fuel rod and the response of submergence, and in addition, the employment of the aforesaid incident angle makes it hard to receive reflected waves from the surface of the fuel rod which is great in terms of energy to facilitate discrimination of waves responsive to submergence. (Kawakami, Y.)

Control rod position detection device

International Nuclear Information System (INIS)

Akita, Haruo; Ogiwara, Sakae.

1996-01-01

The device of the present invention is used in a back-up shut down system of an LMFBR type reactor which is easy for maintenance, has high reliability and can recognize the position of control rods accurately. Namely, a permanent magnet is disposed to a control rod extension tube connected to the lower portion of the control rod. The detector guide tube is disposed in the vicinity of the control rod extension tube. A detector having a detection coil is inserted into a detector tube. With such constitution, the control rod can be detected at one position using the following method. (1) the movement of the magnetic field of the permanent magnet is detected by the detection coil. (2) a plurality of grooves are formed on the control rod extension tube, and the movement of the grooves is detected. In addition, the detection coil is inserted into the detector guide tube, and the signals from the detection coil are inputted to a signal processing circuit disposed at the outside of the reactor vessel using an MI cable to enable the maintenance of the detector. Further, if the detector comprises a detection coil and an excitation coil, the position of a dropped control rod can be recognized at a plurality of points. (I.S.)

Rope wind-up type control rod

International Nuclear Information System (INIS)

Tsuji, Teruaki; Watanabe, Shigeru.

1979-01-01

Purpose: To hold a control rod at a certain position even if the sealed cover of the rod drive mechanism should fail. Constitution: A plurality of friction plates, engaging wheels and a threaded shaft are provided to the wind-up drum for winding up a rope which moves the control rod up and down. While the control rod is adapted to drop by its own weight upon insertion, it is adapted to stop at a predetermined position exactly with no shocks by gradually increasing braking force by the sliding friction caused from the friction plates or the like. A ratch mechanism is provided to the upper portion of the control rod so that the top of the ratch piece may automatically engage the guide passage wall of the control rod upon uncontrolled running of the control rod to prevent further uncontrolled running thereof. (Ikeda, J.)

Analysis of space-time core dynamics on reactor accident at Chernobyl

International Nuclear Information System (INIS)

Takano, Makoto; Shindo, Ryuichi; Yamashita, Kiyonobu; Sawa, Kazuhiro

1987-05-01

Regarding reactor accident at Chernobyl in USSR, core dynamics has been analyzed by COMIC code which solves space-time dependent diffusion equation in three-dimension taking spatial thermohydraulic effect into account. The code was originally developed for high temperature gas-cooled reactors (HTGR), however, has been modified to include light water as coolant, instead of helium, for analysis of the accident. In the analysis, emphasis is placed on spatial effects on core dynamics. The analyses are performed for the cases of modeling the core fully and partially where 6 fuel channels surround one control rod channel. The result shows that the speed of applying void reactivity averaged over the core depends on the power and coolant flow distributions. Therefore, these distributions have potential to influence on the value and the time of peak power estimated by calculation. (author)

Fuel Behaviour at High During RIA and LOCA Accidents; Comportamiento del Combustible de Alto Quemado en Accidents RIA y LOCA

Energy Technology Data Exchange (ETDEWEB)

Barrio del Juanes, M T; Garcia Cuesta, J C; Vallejo Diaz, I; Puebla, Herranz

2001-07-01

Safety analysis of high burnup fuel requires ensuring the acceptable performance under design basis accidents, in particular during conditions representative of Reactivity Accidents (RIA) and Loss-of-Coolant Accidents (LOCA). The report's objective is to compile the state of the art on these issues. This is mainly focused in the effort made to define the applicability of safety criteria to the high burnup fuel. Irradiation damage modifies fuel rod properties, thus the probability of fuel to withstand thermal and mechanical loads during an accident could be quite different compared with unirradiated fuel. From the thermal point of view, fuel conductivity is the most affected property, decreasing notably with irradiation. From the mechanical point of view, a change in the pellet microstructure at its periphery is observed at high burnup (remiffect). Cladding is also effected during operation, showing a significant external and internal corrosion. All these phenomena result in the decrease of efficiency in heat transfer an in the reduction of capability to accommodate mechanical loads; this situation is especially significant at high burnup, when pellet-cladding mechanical interaction is present. Knowledge about these phenomena is not possible without appropriate experimental programmes. The most relevant have been performed in France, Japan, United States and Russia. Results obtained with fuel at high burnup show significant differences with respect to the phenomena observed in fuel at the present discharge burnup. Indeed, this is the encouragement to research about this occurrence. This study is framed within the CSN-CIEMAT agreement, about Fuel Thermo-Mechanical Behaviour at High Burnup. (Author) 172 refs.

Fuel Behaviour at High During RIA and LOCA Accidents; Comportamiento del Combustible de Alto Quemado en Accidents RIA y LOCA

Energy Technology Data Exchange (ETDEWEB)

Barrio del Juanes, M.T.; Garcia Cuesta, J.C.; Vallejo Diaz, I.; Herranz Puebla

2001-07-01

Safety analysis of high burnup fuel requires ensuring the acceptable performance under design basis accidents, in particular during conditions representative of Reactivity Accidents (RIA) and Loss-of-Coolant Accidents (LOCA). The report's objective is to compile the state of the art on these issues. This is mainly focused in the effort made to define the applicability of safety criteria to the high burnup fuel. Irradiation damage modifies fuel rod properties, thus the probability of fuel to withstand thermal and mechanical loads during an accident could be quite different compared with unirradiated fuel. From the thermal point of view, fuel conductivity is the most affected property, decreasing notably with irradiation. From the mechanical point of view, a change in the pellet microstructure at its periphery is observed at high burnup (remiffect). Cladding is also effected during operation, showing a significant external and internal corrosion. All these phenomena result in the decrease of efficiency in heat transfer an in the reduction of capability to accommodate mechanical loads; this situation is especially significant at high burnup, when pellet-cladding mechanical interaction is present. Knowledge about these phenomena is not possible without appropriate experimental programmes. The most relevant have been performed in France, Japan, United States and Russia. Results obtained with fuel at high burnup show significant differences with respect to the phenomena observed in fuel at the present discharge burnup. Indeed, this is the encouragement to research about this occurrence. This study is framed within the CSN-CIEMAT agreement, about Fuel Thermo-Mechanical Behaviour at High Burnup. (Author) 172 refs.

ELECTROMAGNETIC APPARATUS FOR MOVING A ROD

Science.gov (United States)

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.

Individual nuclear fuel rod weighing system

International Nuclear Information System (INIS)

Fogg, J.L.; Smith, J.H.; Vining, G.E.; Howell, C.A.

1985-01-01

An individual nuclear fuel rod weighing system for rods carried on a tray which moves along a materials handling conveyor is discussed. At a first tray position on the conveyor, a lifting device raises the rods off the tray and places them on an overhead ramp. A loading mechanism conveys the rods singly from the overhead ramp onto an overhead scale for individual weighing. When the tray is at a second position on the conveyor, a transfer apparatus transports each weighed rod from the scale back onto the tray

Control rods in LMFBRs: a physics assessment

International Nuclear Information System (INIS)

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 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

Assessment of the potential for high-pressure melt ejection resulting from a Surry station blackout transient

International Nuclear Information System (INIS)

Knudson, D.L.; Dobbe, C.A.

1993-11-01

Containment integrity could be challenged by direct heating associated with a high pressure melt ejection (HPME) of core materials following reactor vessel breach during certain severe accidents. Intentional reactor coolant system (RCS) depressurization, where operators latch pressurizer relief valves open, has been proposed as an accident management strategy to reduce risks by mitigating the severity of HPME. However, decay heat levels, valve capacities, and other plant-specific characteristics determine whether the required operator action will be effective. Without operator action, natural circulation flows could heat ex-vessel RCS pressure boundaries (surge line and hot leg piping, steam generator tubes, etc.) to the point of failure before vessel breach, providing an alternate mechanism for RCS depressurization and HPME mitigation. This report contains an assessment of the potential for HPME during a Surry station blackout transient without operator action and without recovery. The assessment included a detailed transient analysis using the SCDAP/RELAP5/MOD3 computer code to calculate the plant response with and without hot leg countercurrent natural circulation, with and without reactor coolant pump seal leakage, and with variations on selected core damage progression parameters. RCS depressurization-related probabilities were also evaluated, primarily based on the code results

The M5 Fuel Rod Cladding

International Nuclear Information System (INIS)

Mardon, J.P.; Charquet, D.; Senevat, J.

1998-01-01

The large-scale program for the development and irradiation of new Zr alloys started by FRAMATOME and its industrial partners CEZUS and ZIRCOTUBE more than 10 years ago is now enabling FRAGEMA to offer the ternary M5 (ZrNbO) as the cladding material for PWR advanced fuel rods. Compared with the former product (low-tin-Zircaloy-4), this alloy exhibits impressive gains under irradiation at extended burnup (55 GWd/t) relatively to corrosion (factor 3 to 4), hydriding (factor 5 to 6), growth and creep (factor 2 to 3). In this paper, we shall successively address: - the industrial development and manufacturing experience - the corrosion, hydriding, creep and growth performances obtained over a wide range of PWR normal irradiation conditions (France and other countries) up to burnups of 55 GWd/t - The interpretation of these results by means of analytical experiments conducted in test reactors (free growth, creep) and microstructural observations on the irradiated material - and the behaviour under accident (LOCA) and severe environment and irradiation (Li, boiling) conditions. (Author)

Measurements of Burnout Conditions for Flow of Boiling Water in Vertical 3-Rod and 7-Rod Clusters

Energy Technology Data Exchange (ETDEWEB)

Becker, Kurt M; Hernborg, G; Flinta, J E

1964-08-15

The present report deals with measurements of burnout conditions for flow of boiling water in vertical 3-rod and 7-rod clusters. Data were obtained,in respect of heating the rods only, as well as for simultaneous uniform and non-uniform heating of the rods and the shroud. Totally, 520 runs were performed. In the case of equal heat fluxes on all surfaces of the channels, burnout always occurred on the rods, and the data were low by a factor of about 1.3 compared with round duct data. When only the rods were heated, the data showed very low burnout values in comparison with the results for total uniform heating and round ducts. This disagreement was explained by considering the climbing film flow model and the fact that only a fraction of the channel perimeter was heated. For simultaneous and non-uniform heating of the rods and the shroud it was found that the shroud could be overloaded up to 50 per cent without reducing the margin of safety in respect of burnout for the rod cluster. Finally, a correlation for predicting burnout conditions in round ducts, annuli and rod clusters has been presented. This correlation predicts the burnout heat fluxes for the present measurements and previously obtained annuli measurements within {+-} 5 per cent.

Development of a control rod drive

International Nuclear Information System (INIS)

1991-01-01

In the period under review, the computer codes required for transients calculation have been completed, as well as the programs for modelling and testing the hot-gas temperature control by means of combined core rod and reflector rod operation. The specification of requirements to be fulfilled by the rod drive computer and the neutron flux measuring system has been done relying essentially on the data obtained by the transients calculations performed and the resulting informations on operating conditions. The work for optimization of the core rod drive with regard to rod driving speeds and the 'three-point switch' with hysteresis for controlled, automatic core rod operation has been concentrating on the case of specified, normal operation of the reactor. (orig./DG) [de

Self-Assembly of Rod-Coil Block Copolymers

National Research Council Canada - National Science Library

Jenekhe, S

1999-01-01

... the self-assembly of new rod-coil diblock, rod- coil-rod triblock, and coil-rod-coil triblock copolymers from solution and the resulting discrete and periodic mesostmctares with sizes in the 100...

Quivers For Special Fuel Rods-Disposal Of Special Fuel Rods In CASTOR V Casks

International Nuclear Information System (INIS)

Bannani, Amin; Cebula, Wojciech; Buchmuller, Olga; Huggenberg, Roland; Helmut Kuhl

2015-01-01

While GNS casks of the CASTOR family are a suitable means to transfer fuel assemblies (FA) from the NPP to an interim dry storage site, Germanys phase-out of nuclear energy has triggered the demand for an additional solution to dispose of special fuel rods (SFR), normally remaining in the fuel pond until the final shutdown of the NPP. SFR are fuel rods that had to be removed from fuel assemblies mainly due to their special condition, e. g. damages in the cladding of the fuel rods which may have occurred during reactor operations. SFR are usually stored in the spent fuel pond after they are removed from the FA. The quiver for special fuel rods features a robust yet simple design, with a high mechanical stability, a reliable leak-tightness and large safety margins for future requirements on safety analysis. The quiver for special fuel rods can be easily adapted to a large variety of different damaged fuel rods and tailored to the specific need of the customer. The quiver for special fuel rods is adaptable e.g. in length and diameter for use in other types of transport and storage casks and is applicable in other countries as well. The overall concept presented here is a first of its kind solution for the disposal of SFRs via Castor V-casks. This provides an important precondition in achieving the status 'free from nuclear fuel' of the shut down German NPPs

Quivers For Special Fuel Rods-Disposal Of Special Fuel Rods In CASTOR V Casks

Energy Technology Data Exchange (ETDEWEB)

Bannani, Amin; Cebula, Wojciech; Buchmuller, Olga; Huggenberg, Roland [GNS, Essen (Germany); Helmut Kuhl [WTI, Julich (Germany)

2015-05-15

While GNS casks of the CASTOR family are a suitable means to transfer fuel assemblies (FA) from the NPP to an interim dry storage site, Germanys phase-out of nuclear energy has triggered the demand for an additional solution to dispose of special fuel rods (SFR), normally remaining in the fuel pond until the final shutdown of the NPP. SFR are fuel rods that had to be removed from fuel assemblies mainly due to their special condition, e. g. damages in the cladding of the fuel rods which may have occurred during reactor operations. SFR are usually stored in the spent fuel pond after they are removed from the FA. The quiver for special fuel rods features a robust yet simple design, with a high mechanical stability, a reliable leak-tightness and large safety margins for future requirements on safety analysis. The quiver for special fuel rods can be easily adapted to a large variety of different damaged fuel rods and tailored to the specific need of the customer. The quiver for special fuel rods is adaptable e.g. in length and diameter for use in other types of transport and storage casks and is applicable in other countries as well. The overall concept presented here is a first of its kind solution for the disposal of SFRs via Castor V-casks. This provides an important precondition in achieving the status 'free from nuclear fuel' of the shut down German NPPs.

Monitoring device for withdrawing control rods

International Nuclear Information System (INIS)

Higashigawa, Yuichi.

1985-01-01

Purpose: To improve the sensitivity and the responsivity to an equivalent extent to those in the case where local power range monitors are densely arranged near each of the control rods, with no actual but pseudo increase of the number of local power range monitors. Constitution: The monitor arrangement is patterned by utilizing the symmetricity of the reactor core and stored in a monitor designating device. The symmetricity of control rods to be selected and withdrawn by an operator is judged by a control rod symmetry monitoring device, while the symmetricity of the withdrawn control rods is judged by a control rod withdrawal state monitoring device. Then, only when both of the devices judge the symmetricity, the control rods are subjected to gang driving by the control rod drive mechanisms. In this way, monitoring at a high sensitivity and responsivity is enabled with no increase for the number of monitors. (Yoshino, Y.)

Nondestructive assay of HTGR fuel rods

International Nuclear Information System (INIS)

Menlove, H.O.

1974-01-01

Performance characteristics of three different radioactive source NDA systems are compared for the assay of HTGR fuel rods and stacks of rods. These systems include the fast neutron Sb-Be assay system, the 252 Cf ''Shuffler,'' and the thermal neutron PAPAS assay system. Studies have been made to determinethe perturbation on the measurements from particle size, kernel Th/U ratio, thorium content, and hydrogen content. In addition to the total 235 U determination, the pellet-to-pellet or rod-to-rod uniformity of HTGR fuel rod stacks has been measured by counting the delayed gamma rays with a NaI through-hole in the PAPAS system. These measurements showed that rod substitutions can be detected easily in a fuel stack, and that detailed information is available on the loading variations in a uniform stack. Using a 1.0 mg 252 Cf source, assay rates of 2 to 4 rods/s are possible, thus facilitating measurement of 100 percent of a plant's throughput. (U.S.)

Inspection system for Zircaloy clad fuel rods

International Nuclear Information System (INIS)

Yancey, M.E.; Porter, E.H.; Hansen, H.R.

1975-10-01

A description is presented of the design, development, and performance of a remote scanning system for nondestructive examination of fuel rods. Characteristics that are examined include microcracking of fuel rod cladding, fuel-cladding interaction, cladding thickness, fuel rod diameter variation, and fuel rod bowing. Microcracking of both the inner and outer fuel rod surfaces and variations in wall thickness are detected by using a pulsed eddy current technique developed by Argonne National Laboratory (ANL). Fuel rod diameter variation and fuel rod bowing are detected by using two linear variable differential transformers (LVDTs) and a signal conditioning system. The system's mechanical features include variable scanning speeds, a precision indexing system, and a servomechanism to maintain proper probe alignment. Initial results indicate that the system is a very useful mechanism for characterizing irradiated fuel rods

Seismic scrammability of HTTR control rods

International Nuclear Information System (INIS)

Nishiguchi, I.; Iyoku, T.; Ito, N.; Watanabe, Y.; Araki, T.; Katagiri, S.

1990-01-01

Scrammability tests on HTTR (High-Temperature Engineering Test Reactor) control rods under seismic conditions have been carried out and seismic conditions influences on scram time as well as functional integrity were examined. A control rod drive located in a stand-pipe at the top of a reactor vessel, raises and lowers a pair of control rods by suspension cables. Each flexible control rod consists of 10 neutron absorber sections held together by a metal spine passing through the center. It falls into a hole in graphite blocks due to gravity at scram. In the tests, a full scale control rod drive and a pair of control rods were employed with a column of graphite blocks in which holes for rods were formed. Blocks misalignment and contact with the hole surface during earthquakes were considered as major causes of disturbance in scram time. Therefore, the following parameters were set up in the tests: excitation direction, combination or horizontal and vertical excitation, acceleration, frequency and block to block gaps. Main results obtained from tests are as follow. 1) Every scram time obtained under the design conditions was within 6 seconds. On the contrary, the scram times were 5.2 seconds when there were no vibration. Therefore, it was concluded that the seismic effects on scram time were not significant. 2) Scram time became longer with increase in both acceleration and horizontal excitation frequency, and control rods fell very smoothly without any jerkiness. This suggests that collision between control rods and hole surface is the main disturbing factor of falling motion. 3) Mechanical and functional integrity of control rod drive mechanism, control rods and graphite blocks was confirmed after 140 seismic scrammability tests. (author). 10 figs, 1 tab

Control rod testing apparatus

International Nuclear Information System (INIS)

Gaunt, R.R.; Ashman, C.M.

1987-01-01

A control rod testing apparatus is described comprising: a first guide means having a vertical cylindrical opening for grossly guiding a control rod; a second guide means having a vertical cylindrical opening for grossly guiding a control rod. The first and second guide means are supported at axially spaced locations with the openings coaxial; and a substantially cylindrical subassembly having a vertical cylindrical opening therethrough. The subassembly is trapped coaxial with and between the first and second guide means, and the subassembly radially floats with respect to the first and second guide means

ASACUSA's radio-frequency quadrupole decelerator, open to show the four-rod structure along the centre, which crosses 35 resonator chambers formed by the vertical partitions.

CERN Multimedia

Laurent Guiraud

2000-01-01

The Radio-Frequency Quadrupole, RFQD, which further decelerates antiprotons ejected from the Antiproton Decelerator (AD). Starting from a momentum of 100 MeV/c (kinetic energy 5.3 MeV), the RFQD delivers very-low-energy antiprotons, adjustable between 10 and 110 keV, to the experiment ASACUSA. In picture _02, the view from the upstream end shows its 4-rod structure, traversing 35 resonator chambers formed by the vertical partitions. The tank has an inner diameter of 390 mm and is pumped to a vacuum of a few E-8 Torr.

Right Ventricular Ejection Fraction using ECG-Gated First Pass Cardioangiography

Energy Technology Data Exchange (ETDEWEB)

Moon, Young Hee; Lee, Hae Giu; Lee, Sung Yong; Park, Suk Min; Chung, Soo Kyo; Yim, Jeong Ik; Bahk, Yong Whee; Shinn, Kyung Sub; Kim, Young Gyun; Kwon, Soon Seog [Catholic University College of Medicine, Seoul (Korea, Republic of)

1993-03-15

Radionuclide cardioangiography has been widely applied and has played major roles in moninvasive assessment of cardiac function. Three techniques, first-pass gated first and gated equilibrium methods have commonly been used to evaluate right ventricular ejection fraction which usually abnormal in the patients with cardiopulmonary disease. It has been known that the gated first pass method is most accurate method among the three techniques in assessment of fight ventricular ejection fraction. The radionuclide right ventricular ejection fraction values were determined in 13 normal subjects and in 15 patients with chronic obstructive pulmonary disease by the gated first pass method and compared with those of the first pass method because there has been no published data of fight ejection fraction by the gated first pass method were compared with the defects from the pulmonary function test performed in the patients with chronic obstructive pulmomary disease. The results were as follows; 1) The values of fight ventricular ejection fraction by the gated first pass method were 50.1 +- 6.1% in normal subjects and 38.5 +- 8.5 in the patients with chronic obstructive pulmonary disease. There was statistically significant difference between the right ventricular ejection fraction of each of the two groups (p<0.05) 2) The right ventricular ejection fraction by the gated first pass method was not linearly correlated ith FEV{sub 1}, VC. DLCO. and FVC as well as P{sub a}O2 and P{sub a}CO2 of the patients with chronic obstructive pulmonary disease. We concluded that right ventricular ejection fraction by the gated first pass method using radionuclide cardioangiography may be useful in clinical assessment of the right ventricular function.

Development of the automatic control rod operation system for JOYO. Verification of automatic control rod operation guide system

International Nuclear Information System (INIS)

Terakado, Tsuguo; Suzuki, Shinya; Kawai, Masashi; Aoki, Hiroshi; Ohkubo, Toshiyuki

1999-10-01

The automatic control rod operation system was developed to control the JOYO reactor power automatically in all operation modes(critical approach, cooling system heat up, power ascent, power descent), development began in 1989. Prior to applying the system, verification tests of the automatic control rod operation guide system was conducted during 32nd duty cycles of JOYO' from Dec. 1997 to Feb. 1998. The automatic control rod operation guide system consists of the control rod operation guide function and the plant operation guide function. The control rod operation guide function provides information on control rod movement and position, while the plant operation guide function provide guidance for plant operations corresponding to reactor power changes(power ascent or power descent). Control rod insertion or withdrawing are predicted by fuzzy algorithms. (J.P.N.)

Measurements of Burnout Conditions for Flow of Boiling Water in Vertical 3-Rod and 7-Rod Clusters

International Nuclear Information System (INIS)

Becker, Kurt M.; Hernborg, G.; Flinta, J.E.

1964-08-01

The present report deals with measurements of burnout conditions for flow of boiling water in vertical 3-rod and 7-rod clusters. Data were obtained,in respect of heating the rods only, as well as for simultaneous uniform and non-uniform heating of the rods and the shroud. Totally, 520 runs were performed. In the case of equal heat fluxes on all surfaces of the channels, burnout always occurred on the rods, and the data were low by a factor of about 1.3 compared with round duct data. When only the rods were heated, the data showed very low burnout values in comparison with the results for total uniform heating and round ducts. This disagreement was explained by considering the climbing film flow model and the fact that only a fraction of the channel perimeter was heated. For simultaneous and non-uniform heating of the rods and the shroud it was found that the shroud could be overloaded up to 50 per cent without reducing the margin of safety in respect of burnout for the rod cluster. Finally, a correlation for predicting burnout conditions in round ducts, annuli and rod clusters has been presented. This correlation predicts the burnout heat fluxes for the present measurements and previously obtained annuli measurements within ± 5 per cent

Study on the behavior of waterside corroded PWR fuel rods under reactivity initiated accident conditions

International Nuclear Information System (INIS)

Sasajima, Hideo

1989-06-01

One of the highlighted problems from the fuel reliability point of view is a waterside corrosion of fuel cladding which becomes more significant at extended burnup stages. To date, at highly burned fuel, waterside corrosion was recognized as important because cladding oxidation increased with increasing burn-up. In experiments, as the basic research for the study of high burn-up fuel, the test fuel rods were prepressurized to ranges from 3.47 to 3.55 MPa, oxidized artificially to both 10 and 20 μm in thickness. Regarding fabricated oxide thickness of 10 μm, it is corresponded to be transition point from cubic law to linear law as a function of burn-up. Pulse irradiation experiments by NSRR were carried out to study the behavior of waterside corroded PWR type fuels under RIA conditions. Obtained results are: (1) The failure threshold of tested fuels was 110 cal/g·fuel (0.46 KJ/g·fuel) in enthalpy. This showed that the failure threshold of tested fuels was same as that of the past NSRR experimental data. (2) The failure mechanisms of the tested fuel rods was cladding rupture induced by ballooning. No differences in failure mechanisms existed between the past NSRR prepressurized standard fuel and the tested fuels. (3) Cracks were existed without propagating into cladding matrix, so that it was judged that these were not initiation of failure. (4) Whithin this experimental condition, reduction of cladding thickness being attributed to the increase of oxidation did not failure threshold. (author)

A Procedure to Address the Fuel Rod Failures during LB-LOCA Transient in Atucha-2 NPP

Directory of Open Access Journals (Sweden)

Martina Adorni

2011-01-01

Full Text Available Depending on the specific event scenario and on the purpose of the analysis, the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes might be required. This may imply the use of a dedicated fuel rod thermomechanical computer code. This paper provides an outline of the methodology for the analysis of the 2A LB-LOCA accident in Atucha-2 NPP and describes the procedure adopted for the use of the fuel rod thermomechanical code. The methodology implies the application of best estimate thermalhydraulics, neutron physics, and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions are provided by core physics and three-dimensional neutron kinetic coupled thermal-hydraulic system codes calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, which are outside the scope of the current paper.

Mass ejection in failed supernovae: variation with stellar progenitor

Science.gov (United States)

Fernández, Rodrigo; Quataert, Eliot; Kashiyama, Kazumi; Coughlin, Eric R.

2018-05-01

We study the ejection of mass during stellar core-collapse when the stalled shock does not revive and a black hole forms. Neutrino emission during the protoneutron star phase causes a decrease in the gravitational mass of the core, resulting in an outward going sound pulse that steepens into a shock as it travels out through the star. We explore the properties of this mass ejection mechanism over a range of stellar progenitors using spherically symmetric, time-dependent hydrodynamic simulations that treat neutrino mass-loss parametrically and follow the shock propagation over the entire star. We find that all types of stellar progenitor can eject mass through this mechanism. The ejected mass is a decreasing function of the surface gravity of the star, ranging from several M⊙ for red supergiants to ˜0.1 M⊙ for blue supergiants and ˜10-3 M⊙ for Wolf-Rayet stars. We find that the final shock energy at the surface is a decreasing function of the core-compactness, and is ≲ 1047-1048 erg in all cases. In progenitors with a sufficiently large envelope, high core-compactness, or a combination of both, the sound pulse fails to unbind mass. Successful mass ejection is accompanied by significant fallback accretion that can last from hours to years. We predict the properties of shock breakout and thermal plateau emission produced by the ejection of the outer envelope of blue supergiant and Wolf-Rayet progenitors in otherwise failed supernovae.

Reconstitutable control rod spider assembly

International Nuclear Information System (INIS)

Shallenberger, J.M.; Ferian, S.J.

1990-01-01

A reconstitutable control rod/spider assembly includes a hollow connecting finger of the spider having a pair of opposing flat segments formed on the interior thereof and engaging a pair of opposing flat sectors formed on the exterior of a stem extending form the upper end of control rod. The stem also has an externally-threaded portion engaging a nut and a pilot aligning portion for the nut. The nut has a radially flexible and expandable thread-defining element captured in its bore. The segments and sectors allow the rod to be removed and reattached after turning through 180 0 to allow more even wear on the rod. (author)

Estimation of irradiated control rod worth

International Nuclear Information System (INIS)

Varvayanni, M.; Catsaros, N.; Antonopoulos-Domis, M.

2009-01-01

When depleted control rods are planned to be used in new core configurations, their worth has to be accurately predicted in order to deduce key design and safety parameters such as the available shutdown margin. In this work a methodology is suggested for the derivation of the distributed absorbing capacity of a depleted rod, useful in the case that the level of detail that is known about the irradiation history of the control rod does not allow an accurate calculation of the absorber's burnup. The suggested methodology is based on measurements of the rod's worth carried out in the former core configuration and on corresponding calculations based on the original (before first irradiation) absorber concentration. The methodology is formulated for the general case of the multi-group theory; it is successfully tested for the one-group approximation, for a depleted control rod of the Greek Research Reactor, containing five neutron absorbers. The computations reproduce satisfactorily the irradiated rod worth measurements, practically eliminating the discrepancy of the total rod worth, compared to the computations based on the nominal absorber densities.

Two-Stage Dynamics of In Vivo Bacteriophage Genome Ejection

Science.gov (United States)

Chen, Yi-Ju; Wu, David; Gelbart, William; Knobler, Charles M.; Phillips, Rob; Kegel, Willem K.

2018-04-01

Biopolymer translocation is a key step in viral infection processes. The transfer of information-encoding genomes allows viruses to reprogram the cell fate of their hosts. Constituting 96% of all known bacterial viruses [A. Fokine and M. G. Rossmann, Molecular architecture of tailed double-stranded DNA phages, Bacteriophage 4, e28281 (2014)], the tailed bacteriophages deliver their DNA into host cells via an "ejection" process, leaving their protein shells outside of the bacteria; a similar scenario occurs for mammalian viruses like herpes, where the DNA genome is ejected into the nucleus of host cells, while the viral capsid remains bound outside to a nuclear-pore complex. In light of previous experimental measurements of in vivo bacteriophage λ ejection, we analyze here the physical processes that give rise to the observed dynamics. We propose that, after an initial phase driven by self-repulsion of DNA in the capsid, the ejection is driven by anomalous diffusion of phage DNA in the crowded bacterial cytoplasm. We expect that this two-step mechanism is general for phages that operate by pressure-driven ejection, and we discuss predictions of our theory to be tested in future experiments.

CANSWEL-2: a computer model of the creep deformation of Zircaloy cladding under loss-of-coolant accident conditions

International Nuclear Information System (INIS)

Haste, T.J.

1982-07-01

The CANSWEL-2 code models cladding creep deformation under conditions relevant to a loss-of-coolant accident (LOCA) in a pressurised water reactor (PWR). It considers in detail the centre rod of a 3 x 3 nominally square array, taking into account azimuthal non-uniformities in cladding thickness and temperature, and the mechanical restraint imposed on contact with neighbouring rods. Any of the rods in the array may assume a non-circular shape. Models are included for primary and secondary creep, dynamic phase change and superplasticity when both alpha- and beta-phase Zircaloy are present. A simple treatment of oxidation strengthening is incorporated. Account is taken of the anisotropic creep behaviour of alpha-phase Zircaloy which leads to cladding bowing. The CANSWEL-2 model is used both as a stand-alone code and also as part of the LOCA analysis code MABEL-2. (author)

Rod cluster having improved vane configuration

International Nuclear Information System (INIS)

Shockling, L.A.; Francis, T.A.

1989-01-01

This patent describes a pressurized water reactor vessel, the vessel defining a predetermined axial direction of the flow of coolant therewithin and having plural spider assemblies supporting, for vertical movement within the vessel, respective clusters of rods in spaced, parallel axial relationship, parallel to the predetermined axial direction of coolant flow, and a rod guide for each spider assembly and respective cluster of rods. The rod guide having horizontally oriented support plates therewithin, each plate having an interior opening for accommodating axial movement therethrough of the spider assembly and respective cluster of rods. The opening defining plural radially extending channels and corresponding parallel interior wall surfaces of the support plate

Sucker rod motor

Energy Technology Data Exchange (ETDEWEB)

Radzalov, N N; Radzhabov, N A

1983-01-01

The motor consists of rollers mounted on the wellmouth and connected by a flexible rink. Reciprocating mechanism is in the form of a horizontal non-mobile single-side operation cylinder, inside which a plunger and rod are mounted. The working housing of the hydrocylinder is connected to a gas-hydr aulic batter, and when running is connected via plunger to the high pressure source; running in reverse it is connected with a safety valve and automatic control unit. The unit is equipped with a reducer and a mechanical transformer consisting of screw and nut, and which is shutoff with a single-side lining. The plunger rod consists of an auger-like unit. The high pressure source is provided by the injection line of the sucker rod that has been equipped with a reverse valve.

Duke Power Company's control rod wear program

International Nuclear Information System (INIS)

Culp, D.C.; Kitlan, M.S. Jr.

1990-01-01

Recent examinations performed at several foreign and domestic pressurized water reactors have identified significant control rod cladding wear, leading to the conclusion that previously believed control rod lifetimes are not attainable. To monitor control rod performance and reduce safety concerns associated with wear, Duke Power Company has developed a comprehensive control rod wear program for Ag-In-Cd and boron carbide (B 4 C) rods at the McGuire and Catawba nuclear stations. Duke Power currently uses the Westinghouse 17 x 17 Ag-In-Cd control rod design at McGuire Unit 1 and the Westinghouse 17 x 17 hybrid B 4 C control rod design with a Ag-In-Cd tip at McGuire Unit 2 and Catawba Units 1 and 2. The designs are similar, with the exception of the absorber material and clad thickness. There are 53 control rods per unit

Fuel rod pellet loading head

International Nuclear Information System (INIS)

Howell, T.E.

1975-01-01

An assembly for loading nuclear fuel pellets into a fuel rod comprising a loading head for feeding pellets into the open end of the rod is described. The pellets rest in a perforated substantially V-shaped seat through which air may be drawn for removal of chips and dust. The rod is held in place in an adjustable notched locator which permits alignment with the pellets

Conception of a model for the description of the rewetting phase of reactor fuel pins following a loss-of-coolant accident

International Nuclear Information System (INIS)

Hinderer, B.; Schuetzle, R.

1976-10-01

The aim of the present paper has been the development of a model describing rewetting of fuel rods in the reflood phase after a loss of coolant accident of a reactor. Because a suitable solution to the problem could not be found an appropriate model has been implemented into an IKE computer program for transient, two-dimensional heat conductance for a cylindrical rod. Developing this model experimental results of up-to-date literature were used. Remarkable is that very small meshes are necessary around the rewetting front to calculate the rewetting velocity which is strongly dependent on the quench temperature. (orig.) [de