Development of a digital reactivity meter for criticality prediction and control rod worth evaluation in pressurized water reactors

International Nuclear Information System (INIS)

Kuramoto, Renato Y.R.; Miranda, Anselmo F.; Valladares, Gastao Lommez; Prado, Adelk C.

2009-01-01

In this work, we have proposed the development of a digital reactivity meter in order to monitor subcriticality continuously during criticality approach in a PWR. A subcritical reactivity meter can provide an easy prediction of the estimated critical point prior to reactor criticality, without complicated hand calculation. Moreover, in order to reduce the interval of the Physics Tests from the economical point of view, a subcritical reactivity meter can evaluate the control rod worth from direct subcriticality measurement. In other words, count rate of Source Range (SR) detector recorded during the criticality approach could be used for subcriticality evaluation or control rod worth evaluation. Basically, a digital reactivity meter is based on the inverse solution of the kinetic equations of a reactor with the external neutron source in one-point reactor model. There are some difficulties in the direct application of a digital reactivity meter to the subcriticality measurement. When the Inverse Kinetic method is applied to a sufficiently high power level or to a core without an external neutron source, the neutron source term may be neglected. When applied to a lower power level or in the sub-critical domain, however, the source effects must be taken in account. Furthermore, some treatments are needed in using the count rate of Source Range (SR) detector as input signal to the digital reactivity meter. To overcome these difficulties, we have proposed a digital reactivity meter combined with a methodology of the modified Neutron Source Multiplication (NSM) method with correction factors for subcriticality measurements in PWR. (author)

Calculational model based on influence function method for power distribution and control rod worth in fast reactors

International Nuclear Information System (INIS)

Toshio, S.; Kazuo, A.

1983-01-01

A model for calculating the power distribution and the control rod worth in fast reactors has been developed. This model is based on the influence function method. The characteristics of the model are as follows: 1. Influence functions for any changes in the control rod insertion ratio are expressed by using an influence function for an appropriate control rod insertion in order to reduce the computer memory size required for the method. 2. A control rod worth is calculated on the basis of a one-group approximation in which cross sections are generated by bilinear (flux-adjoint) weighting, not the usual flux weighting, in order to reduce the collapse error. 3. An effective neutron multiplication factor is calculated by adjoint weighting in order to reduce the effect of the error in the one-group flux distribution. The results obtained in numerical examinations of a prototype fast reactor indicate that this method is suitable for on-line core performance evaluation because of a short computing time and a small memory size

Analysis of subcritical control rod worth measurements in assembly BZB/3

International Nuclear Information System (INIS)

Giese, H.

1981-07-01

A series of subcritical absorber array measurements was performed in version three of the BIZET assembly BZB in order to check the ability of standard reactor computational codes used by the BIZET participants in predicting control rod worths in large fast reactors. Assembly BZB/3 was a two-zone core with a diameter of about 2.5 m and a core height of 0.89 m, fuelled with plutonium. Fifteen control rod positions and twelve secondary shutdown rod positions were simulated in the core. The measurements comprised the insertion of single absorbers as well as various groups of absorbers and were based on the modified source multiplication method. The KfK analysis was confined to the calculation of eigenvalues for different absorber arrays, also with a view to a comparison with the results of a former BZA evaluation with calculation-to-experiment values of up to C/E ∼ 1.10. The C/E-values found for BZB/3 ranged from 1.02 to 1.10 and did not show a systematic variation at different radial positions or different degrees of absorber asymmetry

Comparison of the worth of control and protection system rods of different design on the basis of the measurements in BN-600 reactor

International Nuclear Information System (INIS)

Vasilyev, B.A.; Roslyakov, V.F.; Farakshin, M.R.

1988-01-01

The results of the worth measurements of the basic and experimental absorbing rods of BN-600 reactor are presented. The procedure used for the rods worth comparison on the basis of calculated and experimental data interpretation is described here. Basic and experimental rods relative worth is also presented. (author). 5 refs, 3 figs, 2 tabs

Analysis of RSG-GAS Control Rod Worth Due to Perturbation Reactivity

International Nuclear Information System (INIS)

Taswanda Taryo

2004-01-01

The control rod interaction effect of RSG-GAS typical working core was studied using a method based on the exact perturbation theory with three simplifying assumptions which require only N+1 criticality calculations. The interaction effect between two interacting rods reached up to 19 % while the interaction effect of multiple interacting rods reached up to 32 % for all (8) control rods involved. The accuracy of the adopted method were extensively investigated to determine the error sources and the magnitude of the error. Through comparison of the present results with ones of the simple summation method, it was obvious that the adopted method was superior in that a significant improvement on the accuracy of the calculated reactivity worth can be achieved with a small number of criticality calculations. (author)

Investigation of control rod worth and nuclear end of life of BWR control rods

International Nuclear Information System (INIS)

Magnusson, Per

2008-01-01

This work has investigated the Control Rod Worth (CRW) and Nuclear End of Life (NEOL) values for BWR control rods. A study of how different parameters affect NEOL was performed with the transport code PHOENIX4. It was found that NEOL, expressed in terms of 10 B depletion, can be generalized beyond the conditions for which the rod is depleted, such as different power densities and void fractions, the corresponding variation in the NEOL will be about 0.2-0.4% 10 B. It was also found that NEOL results for different fuel types and different fuel enrichments have a variation of about 2-3% in 10 B depletion. A comparative study on NHOL and CRW was made between PHOENIX4 and the stochastic Monte Carlo code MCNP. It was found that there is a significant difference, both due to differences in the codes and to limitations in the geometrical modeling in PHOENIX4. Since MCNP is considered more physically correct, a methodology was developed to calculate the nuclear end of life of BWR control rods with MCNP. The advantages of the methodology are that it does not require other codes to perform the depletion of the absorber material, it can describe control rods of any design and it can deplete the control rod absorber material without burning the fuel. The disadvantage of the method is that is it time-consuming

Polynomial curve fitting for control rod worth using least square numerical analysis

International Nuclear Information System (INIS)

Muhammad Husamuddin Abdul Khalil; Mark Dennis Usang; Julia Abdul Karim; Mohd Amin Sharifuldin Salleh

2012-01-01

RTP must have sufficient excess reactivity to compensate the negative reactivity feedback effects such as those caused by the fuel temperature and power defects of reactivity, fuel burn-up and to allow full power operation for predetermined period of time. To compensate this excess reactivity, it is necessary to introduce an amount of negative reactivity by adjusting or controlling the control rods at will. Control rod worth depends largely upon the value of the neutron flux at the location of the rod and reflected by a polynomial curve. Purpose of this paper is to rule out the polynomial curve fitting using least square numerical techniques via MATLAB compatible language. (author)

Intercomparison of rod-worth measurement techniques in a LEU-HTR assembly

International Nuclear Information System (INIS)

Williams, T.; Chawla, R.

1994-01-01

The measurement of absorber-rod worths in the radial reflector of a LEU-HTR pebble bed system is described. Particular emphasis is placed on the choice of complementary measurement techniques to ensure that sensitivities to systematic errors in the calculated parameters used in the analysis are minimised. (author) 3 figs., 3 tabs., 8 refs

Development of external coupling for calculation of the control rod worth in terms of burn-up for a WWER-1000 nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Noori-Kalkhoran, Omid, E-mail: o_noori@yahoo.com [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Yarizadeh-Beneh, Mehdi [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of); Ahangari, Rohollah [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of)

2016-08-15

Highlights: • Calculation of control rod worth in term of burn-up. • Calculation of differential and integral control rod worth. • Developing an external couple. • Modification of thermal-hydraulic profiles in calculations. - Abstract: One of the main problems relating to operation of a nuclear reactor is its safety and controlling system. The most widely used control systems for thermal reactors are neutron absorbent rods. In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 nuclear reactor. External coupling of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of the core in various days. PARCS V2.7 has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. An external coupling algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and control rod worth for different control rod groups have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective Full Power Days (EFPDs) in some steps. Results have been compared with the results of Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR). The results show a good agreement and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.

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)

Implementation of CTRLPOS, a VENTURE module for control rod position criticality searches, control rod worth curve calculations, and general criticality searches

Energy Technology Data Exchange (ETDEWEB)

Smith, L.A.; Renier, J.P.

1994-06-01

A module in the VENTURE reactor analysis code system, CTRLPOS, is developed to position control rods and perform control rod position criticality searches. The module is variably dimensioned so that calculations can be performed with any number of control rod banks each having any number of control rods. CTRLPOS can also calculate control rod worth curves for a single control rod or a bank of control rods. Control rod depletion can be calculated to provide radiation source terms. These radiation source terms can be used to predict radiation doses to personnel and estimate the shielding and long-term storage requirements for spent control rods. All of these operations are completely automated. The numerous features of the module are discussed in detail. The necessary input data for the CTRLPOS module is explained. Several sample problems are presented to show the flexibility of the module. The results presented with the sample problems show that the CTRLPOS module is a powerful tool which allows a wide variety of calculations to be easily performed.

Investigating The Integral Control Rod Worth Of The Miniature Neutron Source Reactor MNSR

International Nuclear Information System (INIS)

Nguyen Hoang Hai; Do Quang Binh

2011-01-01

Determining control rod characteristics is an essential problem of nuclear reactor analysis. In this research, the integral control rod worth of the miniature neutron source reactor MNSR is investigated. Some other parameters of the nuclear reactor, such as core excess reactivity, shut down margin, are also calculated. Group constants for all reactor components are generated by the WIMSD code and then are used in the CITATION code to solve the neutron diffusion equations. The maximum relative error of the calculated results compared with the measurement data is about 3.5%. (author)

POWER LEVEL EFFECT IN A PWR ROD EJECTION ACCIDENT

International Nuclear Information System (INIS)

Diamond, D.J.; Bromley, B.P.; Aronson, A.L.

2002-01-01

The purpose of this study is to determine the effect of the initial power level during a rod ejection accident (REA) on the ejected rod worth and the resulting energy deposition in the fuel. The model used is for the hot zero power (HZP) conditions at the end of a typical fuel cycle for the Three Mile Island Unit 1 pressurized water reactor. PARCS , a transient, three-dimensional, two-group neutron nodal diffusion code, coupled with its own thermal-hydraulics model, is used to perform both steady-state and transient simulations. The worth of an ejected control rod is affected by both power level, and the positions of control banks. As the power level is increased, the worth of a single central control rod tends to drop due to thermal-hydraulic feedback and control bank removal, both of which flatten the radial neutron flux and power distributions. Although the peak fuel pellet enthalpy rise during an REA will be greater for a given ejected rod worth at elevated initial power levels, it is more likely the HZP condition will cause a greater net energy deposition because an ejected rod will have the highest worth at HZP. Thus, the HZP condition can be considered the most conservative in a safety evaluation

Development of a parallel processing couple for calculations of control rod worth in terms of burn-up in a WWER-1000 reactor

Energy Technology Data Exchange (ETDEWEB)

Noori-Kalkhoran, Omid; Ahangari, R. [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor Research school; Shirani, A.S. [Shahid Beheshti Univ., Tehran (Iran, Islamic Republic of). Faculty of Engineering

2017-03-15

In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 reactor. Parallel processing of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of core at different days. PARCS V2.7?has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. A Parallel processing algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and Control rod worth have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective full Power Days (EFPDs) in some steps. Results have been compared with Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR) results. The results show great similarity and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.

The development of the measurement technique of the control rod worth with the inverse kinetics method considering the influence of the steady neutron source

International Nuclear Information System (INIS)

Takeuchi, Mitsuo; Wada, Shigeru; Takahashi, Hiroyuki; Hayashi, Kazuhiko; Murayama, Yoji

2000-09-01

At the research reactor such as JRR-3M, the operation management is carried out in order to ensure safe operation, for example, the excess reactivity is measured regularly and confirmed that it satisfies a safety condition. The excess reactivity is calculated using control rod position in criticality and control rod worth measured by a positive period method (P.P method), the conventional inverse kinetic method (IK method) and so on. The neutron source, however, influences measurement results and brings in a measurement error. A new IK method considering the influence of the steady neutron sources is proposed and applied to the JRR-3M. This report shows that the proposed IK method measures control rod worth more precisely than a conventional IK method. (author)

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

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)

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)

Depletion calculations of adjuster rods in Darlington

Energy Technology Data Exchange (ETDEWEB)

Arsenault, B.; Tsang, K., E-mail: benoit.arsenault@amecfw.com, E-mail: kwok.tsang@amecfw.com [AMEC Foster Wheeler, Toronto, ON (Canada)

2015-07-01

This paper describes the simulation methodology and reactivity worth calculated for aged adjuster rods in the Darlington core. ORIGEN-S IST was applied to simulate the isotope transmutation process of the stainless steel and titanium adjusters. The compositions were used in DRAGON-IST to calculate the change in incremental properties of aged adjusters. Pre-simulations of the reactivity worth of the stainless steel and titanium adjusters in Darlington were performed using RFSP-IST and the results showed that the titanium adjuster rods exhibit faster reactivity-worth drop than that of stainless steel rods. (author)

Validation of the MC{sup 2}-3/DIF3D Code System for Control Rod Worth via the BFS-75-1 Reactor Physics Experiment

Energy Technology Data Exchange (ETDEWEB)

Yun, Sunghwan; Kim, Sang Ji [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In this paper, control rod worths of the BFS-75-1 reactor physics experiments were examined using continuous energy MCNP models and deterministic MC2-3/DIF3D models based on the ENDF/B-VII.0 library. We can conclude that the ENDF/B-VII.0 library shows very good agreement in small-size metal uranium fuel loaded core which is surrounded by the depleted uranium blanket. However, the control rod heterogeneity effect reported by the reference is not significant in this problem because the tested control rod models were configured by single rod. Hence comparison with other control rod worth measurements data such as the BFS-109-2A reactor physics experiment is planned as a future study. The BFS-75-1 critical experiment was carried out in the BFS-1 facility of IPPE in Russia within the framework of validating an early phase of KALIMER- 150 design. The Monte-Carlo model of the BFS- 75-1 critical experiment had been developed. However, due to incomplete information for the BFS- 75-1 experiments, Monte-Carlo models had been generated for the reference criticality and sodium void reactivity measurements with disk-wise homogeneous model. Recently, KAERI performed another physics experiment, BFS-109-2A, by collaborating with Russian IPPE. During the review process of the experimental report of the BFS-109-2A critical experiments, valuable information for the BFS-1 facility which can also be used for the BFS-75-1 experiments was discovered.

Control device for the withdrawal of control rod

International Nuclear Information System (INIS)

Ando, Masaki.

1985-01-01

Purpose: To significantly suppress the maximum value of the control-rod worth upon control rod withdrawal. Constitution: At first, a signal for designating the first class is sent from a class-control section to the group-control section. In the group-control section, the peripheral group among the first class is designated by which the withdrawal of the control rods other than the peripheral group is inhibited and the control-rods in the peripheral group are withdrawn one by one. When all of them have been withdrawn, the group-control section designates the central group of the first class. All the control rods of the central group have been withdrawn, then the group-control section designates the peripheral group of the second class. Thereafter, the central group in the second class is designated. The control rods are thus withdrawn in the same manner hereinafter. The maximum value for the control-rod worth can be decreased by such a withdrawing sequence for the control rods. (Horiuchi, T.)

Control rod interaction models for use in 2D and 3D reactor geometries

International Nuclear Information System (INIS)

Bannerman, R.C.

1985-11-01

Control rod interaction models are developed for use in two-dimensional and three-dimensional reactor geometries. These models allow the total worth of any combination of control rods inserted to be determined from the individual worths in conjunction with an algorithm representing interaction effects between them. The validity of the assumptions is demonstrated for fast and thermal systems showing modelling errors of 1#percent# or less in inserted control rod worths. The models are ideally suited for most reactor systems. (UK)

Rapid and accurate control rod calibration measurement and analysis

International Nuclear Information System (INIS)

Nelson, George W.; Doane, Harry J.

1990-01-01

In order to reduce the time needed to perform control rod calibrations and improve the accuracy of the results, a technique for a measurement, analysis, and tabulation of integral rod worths has been developed. A single series of critical rod positions are determined at constant low power to reduce the waiting time between positive period measurements and still assure true stable reactor period data. Reactivity values from positive period measurements and control rod drop measurements are used as input data for a non-linear fit to the expected control rod integral worth shape. With this method, two control rods can be calibrated in about two hours, and integral and differential calibration tables for operator use are printed almost immediately. Listings of the BASIC computer programs for the non-linear fitting and calibration table preparation are provided. (author)

Evaluation of the rod ejection accident in Westinghouse Pressurized Water Reactors using spatial kinetics methods

International Nuclear Information System (INIS)

Risher, D.H. Jr.

1975-01-01

The consequences of a rod ejection accident are investigated in relation to the latest, high power density Westinghouse reactors. Limiting criteria are presented, based on experimental evidence, and if not exceeded these criteria will ensure that there will be no interference with core cooling capability, and radiation releases, if any, will be within the guidelines of 10CFR100. A basis is presented for the conservative selection of plant parameters to be used in the analysis, such that the analysis is applicable to a wide range of past, present, and future reactors. The calculational method employs a one-dimensional spatial kinetics computer code and a transient fuel heat transfer computer code to determine the hot spot fuel temperature versus time following a rod ejection. Using these computer codes, the most limiting hot channel factor (which does not cause the fuel damage limit criteria to be exceeded) has been determined as a function of the ejected rod worth. By this means, the limit criteria have been translated into ejected rod worths and hot channel factors which can be used effectively by the nuclear designer and safety analyst. The calculational method is shown to be conservative, compared to the results of a three-dimensional spatial kinetics analysis

Experimental Breeder Reactor-II automatic control-rod-drive system

International Nuclear Information System (INIS)

Christensen, L.J.

1983-01-01

A computer-controlled automatic control rod drive system (ACRDS) was designed and operated in EBR-II during reactor runs 121 and 122. The ACRDS was operated in a checkout mode during run 121 using a low worth control rod. During run 122 a high worth control rod was used to perform overpower transient tests as part of the LMFBR oxide fuels transient testing program. The testing program required an increase in power of 4 MW/s, a hold time of 12 minutes and a power decrease of 4 MW/s. During run 122, 13 power transients were performed

Analysis of radially heterogeneous ZPPR-13A benchmark for investigating the spatial dependence of the calculated-to-experiment ratio for control rod worths

International Nuclear Information System (INIS)

Mahalakshmi, B.; Mohanakrishnan, P.

1993-01-01

Investigation were performed on the ZPPR-13A critical assembly to determine the cause of the radial variation of the calculated-to-experimental (C/E) ratio for control rod worth in large heterogeneous cores. The effects of errors in cross section, mesh size, group condensation, transport, and modeling were studied by studied by using two- and three-dimensional diffusion calculations and three-dimensional transport calculations. In that process, the cross-section set and the calculation scheme that are being used for fast reactor design in India have been revalidated. The cross-section set was found to yield satisfactory results. Three-dimensional calculations with adjusted and unadjusted cross sections confirmed that the error in cross sections was largely responsible for the radial dependence of the C/E ratios. The contributions from group condensation and mesh size errors were < 2%, and from modeling errors and transport correction, < 1%. The effect of these errors is insignificant when compared with the effect of the cross-section error. The analysis also showed that even without the adjustment in diffusion coefficient suggested in earlier studies, a satisfactory prediction is found, at least for this benchmark. The diffusion-to-transport correction for control rod worth was found to be -7%

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.)

Analysis of addition of the safety rods at RSG-GAS core

International Nuclear Information System (INIS)

S, Tukiran; S, Tagor Malem; K, Iman

2002-01-01

The silicide fuel loading of the RSG-GAS core is planned to increase from 250 gU to 300 gU. Increasing of fuel loading will prolong the operation cycle length from 25 days to 32,5 days, but ability of reactivity compensation by control rods system decreased because the reactivity shut-down margin is available only 1,03 %, expectation is 2.2 %. One of solutions is added two safety control rods in B-3 and G-10 positions the aim of installing two safety rods (BKP) in RSG-GAS core is to increase core safety margin. So before using the safety control rods in the RSG-GAS core, it is necessary to know its performance, one of the tests showing its performance is to measure the reactivity of the safety control rods. Measurement of safety control rods were done to know each reactivity worth of safety control rods at middle cycle so that the safety rod be used in the RSG-GAS core. Measurement done by using calibration control rods with couple compensation method which always using in the RSG-GAS core to measure the existing control rods. The results of measurement showed that two safety rods (BKP01 and BKP02) have reactivity worth of 93.5 cent and 87.5 cent, respectively. the total reactivity worth of safety control rods is 1.38%. So the two safety rods can be used to increase safety margin of the RSG-GAS core if the fuel is exchanged to 300 gU of loading

Performance analysis of LMFBR control rods

International Nuclear Information System (INIS)

Pitner, A.L.; Birney, K.R.

1975-01-01

Control rods in the FFTF and LMFBR's will consist of pin bundles of stainless steel-clad boron carbide pellets. In the FFTF reference design, sixty-one pins of 0.474-inch diameter each containing a 36-inch stack of 0.362-inch diameter boron carbide pellets comprise a control rod. Reactivity control is provided by the 10 B (n,α) 7 Li reaction in the boron carbide. This reaction is accompanied by an energy release of 2.8 MeV, and heating from this reaction typically approaches 100 watts/cm 3 for natural boron carbide pellets in an LMFBR flux. Performance analysis of LMFBR control rods must include an assessment of the thermal performance of control pins. In addition, irradiation performance with regard to helium release, pellet swelling, and reactivity worth depletion as a function of service time must be evaluated

Calculational benchmark comparisons for a low sodium void worth actinide burner core design

International Nuclear Information System (INIS)

Hill, R.N.; Kawashima, M.; Arie, K.; Suzuki, M.

1992-01-01

Recently, a number of low void worth core designs with non-conventional core geometries have been proposed. Since these designs lack a good experimental and computational database, benchmark calculations are useful for the identification of possible biases in performance characteristics predictions. In this paper, a simplified benchmark model of a metal fueled, low void worth actinide burner design is detailed; and two independent neutronic performance evaluations are compared. Calculated performance characteristics are evaluated for three spatially uniform compositions (fresh uranium/plutonium, batch-averaged uranium/transuranic, and batch-averaged uranium/transuranic with fission products) and a regional depleted distribution obtained from a benchmark depletion calculation. For each core composition, the flooded and voided multiplication factor, power peaking factor, sodium void worth (and its components), flooded Doppler coefficient and control rod worth predictions are compared. In addition, the burnup swing, average discharge burnup, peak linear power, and fresh fuel enrichment are calculated for the depletion case. In general, remarkably good agreement is observed between the evaluations. The most significant difference is predicted performance characteristics is a 0.3--0.5% Δk/(kk) bias in the sodium void worth. Significant differences in the transmutation rate of higher actinides are also observed; however, these differences do not cause discrepancies in the performing predictions

Control-rod parametrical studies in the framework of the PRE-RACINE and RACINE programs

International Nuclear Information System (INIS)

Humbert, G.; Ruelle, B.; Daguzan, G.; Stanculescu, A.; Kappler, F.; Scholtyssek, W.; Bouscavet, D.; Martini, M.; Broccoli, U.

1982-01-01

A control-rod experimental program is presented. This program, established in the frame of PRE-RACINE and RACINE common DEBENE, Italian and French experiments at MASURCA facility, is still under progress at the moment. The results, limited to single central rod worth are already available. For these experiments, a parametrical approach has been used. The effects of rod worth, varied separatly by rod side, boron enrichment and core size, on experiment to calculation relative discrepancy (E-C)/C can be drawn out

Adequacy of the analysis of mock-up control rod experiment with FCA

International Nuclear Information System (INIS)

Mizoo, Nobutatsu; Nakano, Masafumi

1977-07-01

A method of numerical analysis has been investigated for the mock-up control rod experiment of FCA VII-1 assembly constructed as the engineering mock-up of prototype fast breeder reactor MONJU. The results of criticality and B 4 C mock-up control rod worths analysis for the assembly are described in comparison with the experimental ones. The tendency of the C/E value with 10 B enrichment and the interaction effect of the multiple rods array was also examined. Reactivities and the mock-up rods worths were obtained with the X-Y geometry six groups diffusion theory. Twelve kinds of the mock-up rods with different 10 B contents and/or enrichments were used in the experiment; effective cross-sections are provided for each rod by calculation using the collision probability method. Criticality of VII-1 90Z assembly is underestimated for 3 reference critical configurations, ranging from -0.65%Δk/k to -0.77%Δk/k. The C/E values at core center for 12 kinds of B 4 C mock-up rods range from 1.03 to 1.09. The overestimate of the rod worth increases with macroscopic absorption cross-section of the rod region. The C/E values for 24 different arrays of the mock-up rods ranging from single rod to five rods lie between 1.04 and 1.08. The C/E value tends to decrease with increase in the number of rods inserted, the values for five rods arrays being about 4% lower than those for single rod arrays. The calculated interaction effects of the multiple rods arrays are slightly more negative than the experimental ones. (auth.)

Control-rod parametrical studies in the framework of the PRE-RACINE and RACINE programs

International Nuclear Information System (INIS)

Humbert, G.; Ruelle, B.; Daguzan, G.

1982-09-01

The program, established in the frame of PRERACINE and RACINE common DEBENE, Italian and French experiments at MASURCA facility, is still under progress at the moment. The results, limited to single central rod worth are already available. For these experiments, a parametrical approach has been used. The effects of rod worth, varied separatly by rod side, boron enrichment and core size, on experiment to calculation relative discrepancy (E-C)/C can be drawn out

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

Summary of Skoda JS rod drop measurements analysis

International Nuclear Information System (INIS)

Svarny, J.; Krysl, V.

1999-01-01

A summary is presented of the Skoda JS rod drop reactivity measurements analysis provided during last two years based on control rod worth measurements by the outer ion chambers. Standard analysis based on comparisons of dynamics macrocode MOBY-DICK-SK and experimental data is extended to the 8-th group delayed neutron structure and new features of rod drop process are investigated. (author)

KfK analysis of the SUPER-PHENIX-1 control rod experiments. Pt. 1

International Nuclear Information System (INIS)

Giese, H.

1991-03-01

As proposed by the SPX-1 analysis task force, MSM (modified source multiplication) correction factors have been produced for a series of control rod configurations established in the first critical core C1D with minimum fissile loading and in the fully loaded core CMP. The report gives a complete description of the method used at KfK to produce these correction factors and summarises the evaluated experimental results obtained. The KfK method is characterized by a 'two-step-adjustment': A basic reactivity scale adjustment and a subsequent rod worth adjustment. The first adjustment was achieved by 'tuning' either the axialbuckling in the leakage term D B 2 or the average number of neutrons per fission in the production term so that the excess reactivity of the so-called 'Follower-core' with all control rods fully raised was properly reproduced. As this excess reactivity could not be directly determined by an experiment, it had to be assessed from the shut-down worth of the main control system in combination with measured fractions of the S-curve of this system. In the second adjustment, the absorber cross sections were tuned to reproduce experimental rod worths. While for the analysis of the C1D experiments, MSM correction factor calculations were performed in 2D centre-plane geometry only, the analysis of the CMP measurements employed both 2D and 3D calculations. (orig./HP)

Experimental critical loadings and control rod worths in LWR-PROTEUS configurations compared with MCNPX results

International Nuclear Information System (INIS)

Plaschy, M.; Murphy, M.; Jatuff, F.; Seiler, R.; Chawla, R.

2006-01-01

The PROTEUS research reactor at the Paul Scherrer Inst. (PSI) has been operating since the sixties and has already permitted, due to its high flexibility, investigation of a large range of very different nuclear systems. Currently, the ongoing experimental programme is called LWR-PROTEUS. This programme was started in 1997 and concerns large-scale investigations of advanced light water reactors (LWR) fuels. Until now, the different LWR-PROTEUS phases have permitted to study more than fifteen different configurations, each of them having to be demonstrated to be operationally safe, in particular, for the Swiss safety authorities. In this context, recent developments of the PSI computer capabilities have made possible the use of full-scale SD-heterogeneous MCNPX models to calculate accurately different safety related parameters (e.g. the critical driver loading and the shutdown rod worth). The current paper presents the MCNPX predictions of these operational characteristics for seven different LWR-PROTEUS configurations using a large number of nuclear data libraries. More specifically, this significant benchmarking exercise is based on the ENDF/B6v2, ENDF/B6v8, JEF2.2, JEFF3.0, JENDL3.2, and JENDL3.3 libraries. The results highlight certain library specific trends in the prediction of the multiplication factor k eff (e.g. the systematically larger reactivity calculated with JEF2.2 and the smaller reactivity associated with JEFF3.0). They also confirm the satisfactory determination of reactivity variations by all calculational schemes, for instance, due to the introduction of a safety rod pair, these calculations having been compared with experiments. (authors)

Analysis of sodium-void-worths in ZPPR-3 modified phase 3 core

Energy Technology Data Exchange (ETDEWEB)

Takeda, T.; Arai, K.; Otake, I. [Osaka Univ. (JP)

1980-09-15

The sodium-void-worths in the ZPPR-3 modified phase 3 core, in which singularities such as control-rods and sodium-followers were voided, have been analyzed using a unified diffusion coefficient. The unified diffusion coefficient is obtained by applying the Benoist formula to a super-cell consisting of different drawers, and is applicable not only to fuel drawers but also to control-rod drawers or sodium-followers. Using the coefficient the interference effect of neutron streaming between different drawers can be taken into account. The applicability of the unified diffusion coefficient to sodium-void-worth calculations has been checked in a slab model and a RZ model. The sodium-void-worths in the ZPPR-3 modified phase 3 core have been analyzed by carrying out 16-group three-dimensional diffusion calculations using the unified diffusion coefficient and the results have been compared with experimental data. The comparison indicates that the unified diffusion coefficient is useful in calculating the sodium-void-worth in a region including sodium-voided singularities.

Optimized Control Rods of the BR2 Reactor

Energy Technology Data Exchange (ETDEWEB)

Kalcheva, Silva; Koonen, E.

2007-09-15

At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

Optimized Control Rods of the BR2 Reactor

International Nuclear Information System (INIS)

Kalcheva, Silva; Koonen, E.

2007-01-01

At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

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.)

Control Rod Reactivity Measurements in the Aagesta Reactor with the Pulsed Neutron Method

Energy Technology Data Exchange (ETDEWEB)

Bjoereus, K

1969-07-01

An extensive series of control rod measurements was made in the Aagesta reactor during the low power experimental period following the first criticality. This report describes the part of these investigations made with the pulsed neutron method, comprising nearly 300 measurements. The main objective was the determination of control rod reactivity worths for different rods and groups of rods, but some supplementary measurements were also made, e.g. a determination of the prompt neutron decay constant for the delayed critical condition and four different cores. The cores consisted of 20, 32, 68, and 140 fuel elements respectively, and measurements were made at room temperature and with the moderator level close to critical for each core, and for the 140-element core also with full moderator height and at the temperatures 140 deg C and 215 deg C. Both fully and partly inserted control rod groups were investigated. The measurements at critical water level give directly the control rod reactivity worths, whereas those with full water height give the shut-down reactivity. A comparison was made between measured reactivity worths for a number of rod groups and those calculated with the HETERO code. The prompt neutron decay constant at delayed criticality {alpha}{sub 0}={beta}/l, for the full core at 215 deg C was found to be 9.60 {+-} 0.30/sec, corresponding to l = 0.76 {+-} 0.02 msec. The shut-down reactivity with 16 coarse control rods in pos. A-D 22, 40-04, 44, 26 is -5% at 25 deg C and -13% at 215 deg C. The relative error is usually around 8% in the reactivity worths, originating mainly from the higher harmonics content in the measured curves.

Verification and uncertainty evaluation of HELIOS/MASTER nuclear design system

Energy Technology Data Exchange (ETDEWEB)

Song, Jae Seung; Kim, J. C.; Cho, B. O. [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-03-01

A nuclear design system HELIOS/MASTER was established and core follow calculations were performed for Yonggwang Unit 1 cycles 1 through 7 and Yonggwang Unit 3 cycles 1 through 2. The accuracy of HELIOS/MASTER system was evaluated by estimations of uncertainties of reactivity and peaking factors and by comparisons of the maximum differences of isothermal temperature coefficient, inverse boron worth and control rod worth with the CASMO-3/MASTER uncertainties. The reactivity uncertainty was estimated by 362 pcm, and the uncertainties of three-dimensional, axially integrated radial, and planar peaking factors were evaluated by 0.048, 0.034, and 0.044 in relative power unit, respectively. The maximum differences of isothermal temperature coefficient, inverse boron worth and control rod worth were within the CASMO-3/MASTER uncertainties. 17 refs., 17 figs., 10 tabs. (Author)

Nuclear reactor with scrammable part length rod

International Nuclear Information System (INIS)

Bevilacqua, F.

1979-01-01

A new part length rod is provided. It may be used to control xenon induced power oscillations but to contribute to shutdown reactivity when a rapid shutdown of the reactor is required. The part length rod consists of a control rod with three regions. The lower control region is a longer weaker active portion separated from an upper stronger shorter poison section by an intermediate section which is a relative non-absorber of neutrons. The combination of the longer weaker control section with the upper high worth poison section permits the part length rod of this to be scrammed into the core when a reactor shutdown is required but also permits the control rod to be used as a tool to control power distribution in both the axial and radial directions during normal operation

HCPWR type fuel elements design with respect to its control rods

International Nuclear Information System (INIS)

Abbate, M.J.; Sbaffoni, M.M.; Patino, N.E.; Torasso, O.

1992-01-01

The high conversion reactors (HCPWR) can improve the nuclear fuel utilization. One of its present problems is the optimization of the control rods' worth and its relationship with the void coefficient. This investigation means, starting from one reference's design, to analyze several possibilities on the number and distribution of the control rods. As main result, one design including 24 control rods in an optimized distribution, is recommended. (author)

Measurements and analyses on reactivity effects of absorber rods in a light-water moderated UO2 lattices

International Nuclear Information System (INIS)

Murakami, Kiyonobu; Miyoshi, Yoshinori; Hirose, Hideyuki; Suzaki, Takenori

1985-03-01

Reactivity effects and reactivity-interference effects of absorber rods were measured with a cylindrical core aiming to obtain bench-marks for verification of the calculational methods. The core consisted of 2.6 w/o enriched UO 2 fuel rods lattice of which water-to-fuel volume ratio was 1.83. In the experiment, the critical water levels were measured changing neutron absorber content of absorber rods and the distance between two absorber rods in the core center. Monte Calro codes KENO-IV and MULTI-KENO were used to calculate reactivity worthes of absorber rods. The calculational results of effective multiplication factors ranged from 0.978 to 0.999 for the 60 cases of critical cores with inserted absorber rods. The calculational results of absorber worthes agreed to the experimental results within twice of the standerd deviation accompanied with the Monte Calro calculation. (author)

Description and characterization of the ACRR's programmable transient rod withdrawal mode

International Nuclear Information System (INIS)

Boldt, K.R.; Sullivan, W.H.; Kefauver, H.L.

1980-01-01

To satisfy experiment needs for Sandia's Advanced Reactor Safety Program, a programmable Transient Rod Withdrawal (TRW) mode has been developed for the Annular Core Research Reactor (ACRR). The programmable mode is a modification of the existing continuous-withdrawal TRW mode and permits speed and direction changes during the pulse sequence. Basically, a TRW operation is similar to a routine pulse operation except that transient rods are mechanically withdrawn rather than pneumatically fired. Being a pulse-type operation, the TRW mode complies with pulse-mode safety system settings. Control system interlocks prevent the pneumatic firing of rods in the TRW mode. The hardware for the programmable TRW mode includes three ACRR transient rods, the ACRR timer, two rod programmers, a minicomputer and a summing circuit for position indication. Each ACRR transient rod is mechanically driven by a stepping motor (rated torque at 4.24 joules) and is capable of a maximum TRW speed of 26.7 centimeters/ second. The maximum reactivity insertion rate is $2.45/second with a transient rod bank worth of $3.00 and $3.47/second with a bank worth of $4.25, which is expected to be installed soon. The ACRR timer is a multifunctional timer used in all operating modes of the reactor. In the programmable TRW mode, the timer starts the rod programmers and drops regulating rods to terminate the operation. Programmed withdrawal capability is provided by one of two rod programmers (a hardwire-based unit and a microprocessor-based unit). The hardwire unit has eight intervals in which speed, direction and distance are selected by switches on the front panel. The microprocessor-based unit has the capability of 64 intervals in which speed, direction, and distance or time can be specified. Programming this unit is accomplished from the front panel or by inputting data from an HP-9845. minicomputer via a digital I/O interface. Self-test programs in the software provide a continual check of an operating

An evaluation of the control rod modelling approach used in VSOP by comparing its results to the experiments performed in the ASTRA critical facility

International Nuclear Information System (INIS)

Reitsma, F.; Naidoo, D.; Karriem, Z.

2002-01-01

The modelling of strong absorber regions in diffusion theory is a well-known problem and many methods have been developed to accommodate the transport effects in diffusion theory. In this work the method of equivalent cross sections is evaluated for the ASTRA critical facility at the Russian Research Centre - Kurchatov Institute in Moscow. The measured reactivity worths of the control rods situated in the side reflector, are compared with the calculated values making use of equivalent diffusion parameters in VSOP. Favourable results were obtained for the control rods positioned within the first ring of reflector blocks with larger errors obtained for control rods positioned further from the core. Furthermore, the use of an equivalent boron concentration to represent the absorber regions was also investigated and shown to be useful if applied correctly and with care. However, the practical difficulties and restrictions imposed by the two approaches make the investigation of an alternative method, which should remove these shortcomings, attractive. (author)

Impact of mesh points number on the accuracy of deterministic calculations of control rods worth for Tehran research reactor

International Nuclear Information System (INIS)

Boustani, Ehsan; Amirkabir University of Technology, Tehran; Khakshournia, Samad

2016-01-01

In this paper two different computational approaches, a deterministic and a stochastic one, were used for calculation of the control rods worth of the Tehran research reactor. For the deterministic approach the MTRPC package composed of the WIMS code and diffusion code CITVAP was used, while for the stochastic one the Monte Carlo code MCNPX was applied. On comparing our results obtained by the Monte Carlo approach and those previously reported in the Safety Analysis Report (SAR) of Tehran research reactor produced by the deterministic approach large discrepancies were seen. To uncover the root cause of these discrepancies, some efforts were made and finally was discerned that the number of spatial mesh points in the deterministic approach was the critical cause of these discrepancies. Therefore, the mesh optimization was performed for different regions of the core such that the results of deterministic approach based on the optimized mesh points have a good agreement with those obtained by the Monte Carlo approach.

Impact of mesh points number on the accuracy of deterministic calculations of control rods worth for Tehran research reactor

Energy Technology Data Exchange (ETDEWEB)

Boustani, Ehsan [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Amirkabir University of Technology, Tehran (Iran, Islamic Republic of). Energy Engineering and Physics Dept.; Khakshournia, Samad [Amirkabir University of Technology, Tehran (Iran, Islamic Republic of). Energy Engineering and Physics Dept.

2016-12-15

In this paper two different computational approaches, a deterministic and a stochastic one, were used for calculation of the control rods worth of the Tehran research reactor. For the deterministic approach the MTRPC package composed of the WIMS code and diffusion code CITVAP was used, while for the stochastic one the Monte Carlo code MCNPX was applied. On comparing our results obtained by the Monte Carlo approach and those previously reported in the Safety Analysis Report (SAR) of Tehran research reactor produced by the deterministic approach large discrepancies were seen. To uncover the root cause of these discrepancies, some efforts were made and finally was discerned that the number of spatial mesh points in the deterministic approach was the critical cause of these discrepancies. Therefore, the mesh optimization was performed for different regions of the core such that the results of deterministic approach based on the optimized mesh points have a good agreement with those obtained by the Monte Carlo approach.

Evaluation of fuel rods behavior - under irradiation test

International Nuclear Information System (INIS)

Lameiras, F.S.; Terra, J.L.; Pinto, L.C.M.; Dias, M.S.; Pinheiro, R.B.

1981-04-01

By the accompanying of the irradiation of instrumented test fuel rods simulating the operational conditions in reactors, plus the results of post - irradiation exams, tests, evaluation and calibration of analitic modelling of such fuel rods is done. (E.G.) [pt

Advances in methods of commercial FBR core characteristics analyses. Investigations of a treatment of the double-heterogeneity and a method to calculate homogenized control rod cross sections

Energy Technology Data Exchange (ETDEWEB)

Sugino, Kazuteru [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center; Iwai, Takehiko

1998-07-01

A standard data base for FBR core nuclear design is under development in order to improve the accuracy of FBR design calculation. As a part of the development, we investigated an improved treatment of double-heterogeneity and a method to calculate homogenized control rod cross sections in a commercial reactor geometry, for the betterment of the analytical accuracy of commercial FBR core characteristics. As an improvement in the treatment of double-heterogeneity, we derived a new method (the direct method) and compared both this and conventional methods with continuous energy Monte-Carlo calculations. In addition, we investigated the applicability of the reaction rate ratio preservation method as a advanced method to calculate homogenized control rod cross sections. The present studies gave the following information: (1) An improved treatment of double-heterogeneity: for criticality the conventional method showed good agreement with Monte-Carlo result within one sigma standard deviation; the direct method was consistent with conventional one. Preliminary evaluation of effects in core characteristics other than criticality showed that the effect of sodium void reactivity (coolant reactivity) due to the double-heterogeneity was large. (2) An advanced method to calculate homogenize control rod cross sections: for control rod worths the reaction rate ratio preservation method agreed with those produced by the calculations with the control rod heterogeneity included in the core geometry; in Monju control rod worth analysis, the present method overestimated control rod worths by 1 to 2% compared with the conventional method, but these differences were caused by more accurate model in the present method and it is considered that this method is more reliable than the conventional one. These two methods investigated in this study can be directly applied to core characteristics other than criticality or control rod worth. Thus it is concluded that these methods will

Physics characteristics of a CANDU-600 with repositional adjuster rods fuelled with MOX or natural uranium

International Nuclear Information System (INIS)

Boczar, P.G.

1985-06-01

Repositioning the adjuster rods in 4 axial banks in future CANDU-600 reactors would permit the flexibility of grading the inner and outer banks to achieve optimal flattening of the axial power distribution for any particular fuel. With the 4 banks identical, acceptable axial power profiles can be achieved for both MOX and natural uranium fuels. Future work is to be directed at assessing the impact of lower zone controller and shutoff rod worth in the configuration of reactivity devices considered here, and if necessary, in identifying means of increasing their worth

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

International Nuclear Information System (INIS)

Waki, Toshikazu; Yamada, Motoyuki; Horikawa, Yoshihiko; Miyake, Yusuke; Sakashita, Akira

2009-01-01

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

Method for operating a nuclear reactor with scrammable part length rod

International Nuclear Information System (INIS)

Bevilacqua, F.

1979-01-01

A new part length rod is provided which may be used to control xenon induced power oscillations but also to contribute to shutdown reactivity when a rapid shutdown of the reactor is required. The part length rod consists of a control rod with three regions. The lower control region is a longer weaker active portion separated from an upper stronger shorter poison section by an intermediate section which is a relative non-absorber of neutrons. The combination of the longer weaker control section with the upper high worth poison section permits the part length rod to be scrammed into the core. When a reactor shutdown is required but also permits the control rod to be used as a tool to control power distribution in both the axial and radial directions during normal operation

Control rod effects with plutonium recycle in a PWR

International Nuclear Information System (INIS)

Nash, G.; Muehl, G.J.; Gibson, I.H.

1979-03-01

A study has been made on a PWR loaded partly and wholly with plutonium to determine the changes in shutdown margin compared with an enriched uranium core. Lattice calculations are used to generate cell constants for core calculations. Three fuel loadings were considered, all uranium, 30% (approximately) of the assemblies plutonium in natural uranium, and all plutonium. The equilibrium fuel management schemes adopted in each case are based on the standard three cycle equal size batch scheme. Detailed calculations of power and irradiation distributions through the cycles have been carried out to provide a starting point for the control rod worth and requirement calculations. Control rod worths are reduced in a plutonium core because of the harder spectrum and higher fuel absorption cross sections. Furthermore, the control rod requirements for shutdown increase because of the increase in fuel and moderator temperature coefficients. This results in a reduction in shutdown margin. The magnitude of these changes is fully analysed in the report. The significance of these reductions depends on the detail of the safety argument but reductions of these sizes are unlikely to be acceptable. The data provided in this report could be used to give a first estimate of the plutonium loading acceptable given the safety assessment of the normal uranium core. (U.K.)

Evaluation of LWR fuel rod behavior under operational transient conditions

International Nuclear Information System (INIS)

Nakamura, M.; Hiramoto, K.; Maru, A.

1984-01-01

To evaluate the effects of fission gas flow and diffusion in the fuel-cladding gap on fuel rod thermal and mechanical behaviors in light water reactor (LWR) fuel rods under operational transient conditions, computer sub-programs which can calculate the gas flow and diffusion have been developed and integrated into the LWR fuel rod performance code BEAF. This integrated code also calculates transient temperature distribution in the fuel-pellet and cladding. The integrated code was applied to an analysis of Inter Ramp Project data, which showed that by taking into account the gas flow and diffusion effects, the calculated cladding damage indices predicted for the failed rods in the ramp test were consistent with iodine-SCC (Stress Corrosion Cracking) failure conditions which were obtained from out-of-reactor pressurized tube experiments with irradiated Zircaloy claddings. This consistency was not seen if the gas flow and diffusion effects were neglected. Evaluation were also made for the BWR 8x8 RJ fuel rod temperatures under power ramp conditions. (orig.)

Two codes used in analysis of rod ejection accident for Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Zhu Xinguan

1987-12-01

Two codes were developed to analyse rod ejection accident for Qinshan Nuclear Power Plant. One was based on point model with temperature reactivity feedback. In this code, the worth of ejected rod was obtained under'adiabatic' approximation. In the other code, the Nodal Green's Function Method was used to solve space-time dependent neutron diffusion equation. Using these codes, the transient core-power have been calculated for two rod ejection cases at beginning of core-life in Qinshan Nuclear Power Plant

KfK-analysis of the SUPER-PHENIX-1 control rod experiments. Pt. 2

International Nuclear Information System (INIS)

Giese, H.

1992-10-01

Using standard data and codes, this first analysis campaign led to results that drastically overestimated measured control rod worths with C/E ratios ranging from 1.17 to 1.28. These findings were in sharp contrast to the results of the majority of earlier control rod experiments in zero-power facilities, where C/E ratios were usually comprised between 1.0 and 1.1. Investigations were then launched to identify the origin of this discrepancy. By comparison with the calculation methods employed by the other working groups engaging in the analysis of the SPX-1 experiments it was found that the principal problem arose from the KfK/BN-procedure used for the homogenization of control rod absorber cross sections. The specific failure of this standard procedure in the case of the SPX-1 analysis is ascribed to the extremely heterogeneous structure of the SPX-1 control rods. In an attempt to improve on this point, a new method for the production of homogenized absorber cross sections was developed and applied to the SPX-1 analysis. The report concludes with a description of this method and a survey of the results obtained for SPX-1. It is found that this revised analysis leads to a significantly improved agreement of measured and calculated control rod worths and to a better consistency with the results of earlier control rod experiments in zero-power facilities. (orig.)

Measurement of reactivity worths of burnable poison rods in enriched uranium graphite-moderated core simulated to high temperature gas cooled reactor

International Nuclear Information System (INIS)

Akino, Fujiyoshi; Takeuchi, Motoyoshi; Kitadate, Kenji; Yoshifuji, Hisashi; Kaneko, Yoshihiko

1980-11-01

As the core design for the Experimental Very High Temperature Gas Cooled Reactor progresses, evaluation of design precision has become increasingly important. For a high precision design, it is required to have adequate group constants based on accurate nuclear data, as well as calculation methods properly describing the physical behavior of neutrons. We, therefore, assembled a simulation core for VHTR, SHE-14, using a graphite-moderated 20%-enriched uranium Semi-Homogeneous Experimental Critical Facility (SHE), and obtained useful experimental data in evaluating the design precision. The VHTR is designed to accommodate burnable poison and control rods for reactivity compensation. Accordingly, the experimental burnable poison rods which are similar to those to be used in the experimental reactor were prepared, and their reactivity values were measured in the SHE-14 core. One to three rods of the above experimental burnable poison rods were inserted into the central column of the SHE-14 core, and the reactivity values were measured by the period and fuel rod substitution method. The results of the measurements have clearly shown that due to the self-shielding effect of B 4 C particles the reactivity value decreases with increasing particle diameter. For the particle diameter, the reactivity value is found to increase linearly with the logarithm of boron content. The measured values and those calculated are found to agree with each other within 5%. These results indicate that the reactivity of the burnable poison rod can be estimated fairly accurately by taking into account the self-shielding effect of B 4 C particles and the heterogeneity of the lattice cell. (author)

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

Evaluation of accuracy of Monte Carlo code MVP with VHTRC experiments. Multiplication factor at criticality, burnable poison worth and void worth

International Nuclear Information System (INIS)

Nojiri, Naoki; Yamashita, Kiyonobu; Fiujimoto, Nozomu; Nakano, Masaaki , Yamane, Tsuyoshi; Akino, Fujiyoshi.

1997-11-01

Experimental data of VHTRC (Very High Temperature Reactor Critical Assembly) were analyzed using Monte Carlo code MVP (general purpose Monte Carlo code of neutron and photon transport calculations based on the continuous energy method). The calculation accuracy of the code was evaluated by the analysis for nuclear characteristics of a HTGR (high temperature gas-cooled reactor). The MVP code can analyze with a detailed three-dimensional core model with a few approximations. The HTGRs have following characteristics from view point of nuclear design : they have burnable poisons, many void holes, namely, the control insertion holes and so on. Taking account of these characteristics, multiplication factor at criticality, burnable poison worth, and void worth were evaluated. The maximum calculation errors were 0.8%Δk, 7%, and 25% respectively, From these results, it can be concluded that the MVP code is able to be applied to the nuclear characteristics analysis of the HTGR like the High Temperature Engineering Test Reactor (HTTR). (author)

Determination of the most reactivity control rod by pseudo-harmonics perturbation method

International Nuclear Information System (INIS)

Freire, Fernando S.; Silva, Fernando C.; Martinez, Aquilino S.

2005-01-01

Frequently it is necessary to compute the change in core multiplication caused by a change in the core temperature or composition. Even when this perturbation is localized, such as a control rod inserted into the core, one does not have to repeat the original criticality calculation, but instead we can use the well-known pseudo-harmonics perturbation method to express the corresponding change in the multiplication factor in terms of the neutron flux expanded in the basis vectors characterizing the unperturbed core. Therefore we may compute the control rod worth to find the most reactivity control rod to calculate the fast shutdown margin. In this thesis we propose a simple and precise method to identify the most reactivity control rod. (author)

Comparable Worth Theory and Policy.

Science.gov (United States)

Wittig, Michele Andrisin; Lowe, Rosemary Hays

1989-01-01

Provides different perspectives on comparable worth issues. Covers the following topics: (1) competing explanations for the wage gap; (2) indirect approaches to wage equity; (3) the need for a direct approach to wage equity; (4) job evaluation; (5) application of comparable worth principles to compensation systems; and (6) strategies for adopting…

Detection of a leaking boron-carbide control rod in a TRIGA Mark I reactor

Energy Technology Data Exchange (ETDEWEB)

Blotcky, A J; Arsenault, L J [General Medical Research, Veterans Administration Hospital, Omaha (United States)

1974-07-01

During a routine quarterly inspection of the boron-carbide control rods of the Omaha Veterans Administration Hospital 18 kW Triga Mark I reactor, a pin hole leak was detected approximately 3 mm from the chamfered edge. The leak was found by observing bubbles when the rod was withdrawn from the reactor tank for visual observation, and could not be seen with the naked eye. This suggests that pin hole leaks could occur and not be visually detected in control rods and fuel elements examined underwater. A review of the rod calibrations showed that the leak had not caused a loss in rod worth. Slides will be presented showing the bubbles observed during the inspection, together with an unmagnified and magnified view of the pin hole. (author)

Detection of a leaking boron-carbide control rod in a TRIGA Mark I reactor

International Nuclear Information System (INIS)

Blotcky, A.J.; Arsenault, L.J.

1974-01-01

During a routine quarterly inspection of the boron-carbide control rods of the Omaha Veterans Administration Hospital 18 kW Triga Mark I reactor, a pin hole leak was detected approximately 3 mm from the chamfered edge. The leak was found by observing bubbles when the rod was withdrawn from the reactor tank for visual observation, and could not be seen with the naked eye. This suggests that pin hole leaks could occur and not be visually detected in control rods and fuel elements examined underwater. A review of the rod calibrations showed that the leak had not caused a loss in rod worth. Slides will be presented showing the bubbles observed during the inspection, together with an unmagnified and magnified view of the pin hole. (author)

Establishment of analysis procedure for control rod reactivity worth

Energy Technology Data Exchange (ETDEWEB)

Song, Hoon; Kim, Young Il; Kim, Sang Ji; Kim, Young In

2001-03-01

As to the calculation method of control rod reactivity relating to hexagonal assembly, which are used generally in fast reactor, we have investigated the calculation method, the problems to rise during calculation, the degrees of calculation and the enhancement of calculation modeling so on, and estimated the application of calculation method through comparison and analysis of calculation result using the effective cross section generation system, TRANSX/TWODANT, and neutron flux calculation system, diffusion theory code DIF-3D, which are belonged to K-CORE System, and determined the basic calculation method, and extracted the present calculation problem in case of application in K-CORE System and the future improvement items so on.

Establishment of analysis procedure for control rod reactivity worth

International Nuclear Information System (INIS)

Song, Hoon; Kim, Young Il; Kim, Sang Ji; Kim, Young In

2001-03-01

As to the calculation method of control rod reactivity relating to hexagonal assembly, which are used generally in fast reactor, we have investigated the calculation method, the problems to rise during calculation, the degrees of calculation and the enhancement of calculation modeling so on, and estimated the application of calculation method through comparison and analysis of calculation result using the effective cross section generation system, TRANSX/TWODANT, and neutron flux calculation system, diffusion theory code DIF-3D, which are belonged to K-CORE System, and determined the basic calculation method, and extracted the present calculation problem in case of application in K-CORE System and the future improvement items so on

Damage analysis of ceramic boron absorber materials in boiling water reactors and initial model for an optimum control rod management

International Nuclear Information System (INIS)

Schulz, W.

2000-01-01

Operating experience has proved so far that BWR control rods cannot be used for the total reactor life time as originally presumed, but instead has to be considered as a consumable article. After only few operating cycles, the mechanism of absorber failure has been shown to be neutron induced boron carbide swelling and stress cracking of the absorber tubes, followed by erosion of the absorber material. In the case that operation of such a control rod is continued in control cells, this can lead to an increase of the local power density distribution in the core and, under certain conditions, can even cause fuel rod damage. A non destructive testing method has been developed called 'UNDERWATER NEUTRON RADIOGRAPHY' applicable for any BWR control rod. 'Lead-control rods' being radiographed are used to evaluate their actual nuclear worth by the help of a special analytical procedure developed and verified by the author. Nuclear worth data plotted against bum up history data will allow to create an 'EMPIRIC MODEL'. This model includes the basic idea of operating control rods of a certain design first in a control position up to a target fluence limited to an amount just below the appearance of control rod washout. Afterwards they have to be moved in a shut down position to work therefor the total remaining holding period. The initial model is applicable to any CR-design as long as sufficient measuring-data and thus data about the nuclear worth are available. The results of these experiences are extrapolated to the whole reactor holding period. After modelling no further measurements of this particular control rod type are necessary in any reactor. The second focal point is to provide an APPROXIMATION EQUATION. By knowing the absorber radius, B 4 C density and absorber enclosure data an engineer will calculate reliably the working life of any control rod design on control position. indicated as maximum allowable neutron fluence margin until absorber wash-out starts. This

Comparison of cell homogenization methods considering interaction effect between fuel cells and control rod cells

International Nuclear Information System (INIS)

Takeda, T.; Uto, N.

1988-01-01

Several methods to determine cell-averaged group cross sections and anisotropic diffusion coefficients which consider the interaction effect between core fuel cells and control rods or control rod followers have been compared to discuss the physical meaning included in cell homogenization. As the cell homogenization methods considered are the commonly used flux-weighting method, the reaction rate preservation method and the reactivity preservation method. These homogenization methods have been applied to control rod worth calculations in 1-D slab cores to investigate their applicability. (author). 6 refs, 2 figs, 9 tabs

Validation of Westinghouse integrated code POLCA-T against OECD NEACRP-L-335 rod ejection benchmark

Energy Technology Data Exchange (ETDEWEB)

Panayotov, Dobromir [Westinghouse Electric Sweden AB, Vaesteraas, SE-721 63 (Sweden)

2008-07-01

This paper describes the work performed and results obtained in the validation of the POLCA-T code against NEACRP PWR Rod Ejection Transients Benchmark. Presented work is a part of the POLCA-T licensing Assessment Data Base for BWR Control Rod Drop Accident (CRDA) Application. The validation against a PWR Rod Ejection Accidents (REA) Benchmark is relevant for the validation of the code for BWR CRDA, as the analyses of both transients require identical phenomena to be modelled. All six benchmark cases have been analyzed in the presented work. Initial state steady-state calculations including boron search, control rod worth, and final state power search have been performed by POLCA7 code. Initial state boron adjustment and steady-state CR worth as well as the transient analyses were performed by POLCA-T code. Benchmark results including 3D transient power distributions are compared with reference PANTHER solutions and published results of other codes. Given the similarity of the kinetics modelling for a BWR CRDA and a PWR REA and the fact that POLCA-T accurately predicts the local transient power and thus, the resulting fuel enthalpy, it is concluded that POLCA-T is a state-of-art tool also for BWR CRDA analysis. (author)

Validation of Westinghouse integrated code POLCA-T against OECD NEACRP-L-335 rod ejection benchmark

International Nuclear Information System (INIS)

Panayotov, Dobromir

2008-01-01

This paper describes the work performed and results obtained in the validation of the POLCA-T code against NEACRP PWR Rod Ejection Transients Benchmark. Presented work is a part of the POLCA-T licensing Assessment Data Base for BWR Control Rod Drop Accident (CRDA) Application. The validation against a PWR Rod Ejection Accidents (REA) Benchmark is relevant for the validation of the code for BWR CRDA, as the analyses of both transients require identical phenomena to be modelled. All six benchmark cases have been analyzed in the presented work. Initial state steady-state calculations including boron search, control rod worth, and final state power search have been performed by POLCA7 code. Initial state boron adjustment and steady-state CR worth as well as the transient analyses were performed by POLCA-T code. Benchmark results including 3D transient power distributions are compared with reference PANTHER solutions and published results of other codes. Given the similarity of the kinetics modelling for a BWR CRDA and a PWR REA and the fact that POLCA-T accurately predicts the local transient power and thus, the resulting fuel enthalpy, it is concluded that POLCA-T is a state-of-art tool also for BWR CRDA analysis. (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

Development of a program for evaluating the temperature of SMART-P fuel rod

Energy Technology Data Exchange (ETDEWEB)

Cheon, Jin Sik; Lee, Byung Ho; Koo, Yang Hyun; Oh, Je Yong; Yim, Jeong Sik; Sohn, Dong Seong

2003-11-01

A code for evaluating the temperature of SMART-P fuel rod has been developed. Finite Element (FE) method is adopted for the developed code sharing the user subroutines which has been prepared for the ABAQUS commercial FE code. The developed program for SMART-P fuel rod corresponds to a nonlinear transient heat transfer problem, and uses a sparse matrix solver for FE equations during iterations at every time step. The verifications of the developed program were conducted using the ABAQUS code. Steady state and transient problems were analyzed for 1/8 rod model due to the symmetry of the fuel rod and full model. From the evaluation of temperature for the 1/8 rod model at steady state, maximal error of 0.18 % was present relative to the ABAQUS result. Analysis for the transient problem using the fuel rod model resulted in the same as the variation of centerline temperature from the ABAQUS code during a hypothetical power transient. Also, given a power depression in fuel meat as a function of burnup, its effect on the centerline temperature was more precisely evaluated by the developed program compared to the ABAQUS code. The distribution of heat flux for the entire cross section and surface was almost identical for the two codes.

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

An Evaluation on the Fluid Elastic Instability of the Fuel Rod for OPR1000 Plants

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyeong Koo; Jeon, Sang Yoon; Lee, Kyu Seok; Kim, Jeong Ha; Lee, Sang Jong [Reactor Core Technology Department, Korea Nuclear Fuel, 493, Deogjin, Yuseong, Daejeon, 305-353 (Korea, Republic of)

2009-06-15

The fuel assembly for a typical PWR (Pressurized Water Reactor) plant suffers severe operating conditions during its lifetime such as high temperature, high pressure and massive coolant passing through the fuel assembly with high speed. Moreover, recently nuclear fuel is requested not only to operate under more severe operation conditions for example high burnup, longer cycle and power up-rate, but also to maintain its integrity in spite of the operation severity. Lots of vendors, therefore, have poured their endeavor to develop an advanced fuel in order to meet these requirements. However, the fuel failures are still reported from time to time. In general, fuel failure mechanisms known as significant causes of PWR fuel failure are grid to rod fretting, corrosion of the cladding, pellet cladding interaction and debris induced fretting. Especially, since the fuel assembly is very tall and flexible structure and the flow velocity of reactor coolant is pretty high, flow induced vibration (FIV) of fuel rod is an inevitable phenomenon in PWR fuel and the energy vibrating fuel rod continually provided by coolant flow can become a root cause of the fuel failure like grid to rod fretting. Moreover, the cross flow of the coolant is highly susceptible to cause the fluid elastic instability (FEI) which produces extraordinarily big amplitudes of the fuel rod suddenly and is eventually ended up fuel failure within very short-term. The FIV problem, therefore, has to be evaluated carefully to avoid unexpected fuel failure. At present, the susceptibility to vibration damage of the fuel rod for OPR1000 plants has been estimated by the comparison of natural frequencies of every fuel rod span with recognized external excitation frequencies like coolant pump blade passing frequencies, vortex shedding frequencies and lower support structure vibration frequencies. That is, in order to prevent fuel failure due to the external excitation, the natural frequencies of unsupported lengths of

Development of a reactivity worth correction scheme for the one-dimensional transient analysis

International Nuclear Information System (INIS)

Cho, J. Y.; Song, J. S.; Joo, H. G.; Kim, H. Y.; Kim, K. S.; Lee, C. C.; Zee, S. Q.

2003-11-01

This work is to develop a reactivity worth correction scheme for the MASTER one-dimensional (1-D) calculation model. The 1-D cross section variations according to the core state in the MASTER input file, which are produced for 1-D calculation performed by the MASTER code, are incorrect in most of all the core states except for exactly the same core state where the variations are produced. Therefore this scheme performs the reactivity worth correction factor calculations before the main 1-D transient calculation, and generates correction factors for boron worth, Doppler and moderator temperature coefficients, and control rod worth, respectively. These correction factors force the one dimensional calculation to generate the same reactivity worths with the 3-dimensional calculation. This scheme is applied to the control bank withdrawal accident of Yonggwang unit 1 cycle 14, and the performance is examined by comparing the 1-D results with the 3-D results. This problem is analyzed by the RETRAN-MASTER consolidated code system. Most of all results of 1-D calculation including the transient power behavior, the peak power and time are very similar with the 3-D results. In the MASTER neutronics computing time, the 1-D calculation including the correction factor calculation requires the negligible time comparing with the 3-D case. Therefore, the reactivity worth correction scheme is concluded to be very good in that it enables the 1-D calculation to produce the very accurate results in a few computing time

Analysis of a control rod ejection transient in a mox-fuelled PWR

International Nuclear Information System (INIS)

Lenain, R.; Mathonniere, G.; Perrutel, J.P.; Schaeffer, H.; Stelletta, S.; Lam Hime, M.

1988-09-01

The decision to use mixed-oxide (MOX) fuel in PWR's involved re-investigation of a certain number of accidents and notably control rod ejection transients. It has thus been shown that this accident would be no more severe than in the case of all-uranium cores, since the positive effects on the ejected rod worth would counterbalance the negative effects on the delayed neutron fraction. A new approach to the kinetics aspect of the calculation method for this accident is also presented, involving a 3-D kinetic calculation with only a few axial meshes

Development of program for evaluating the temperature of Zr-U metallic fuel rod

International Nuclear Information System (INIS)

Chun, J. S.; Lee, B. H.; Ku, Y. H.; Oh, J. Y.; Im, J. S.; Sohn, D. S.

2003-01-01

A code for evaluating the temperature of Zr-U metallic rod has been developed. Finite element (FE) method is adopted for the developed code sharing the user subroutines which has been prepared for the ABAQUS commercial FE code. The developed program for the Zr-U metallic fuel rod corresponds to a nonlinear transient heat transfer problem, and uses a sparse matrix solver for FE equations during iterations at every time step. The verifications of the developed program were conducted using the ABAQUS code. Steady state and transient problems were analyzed for 1/8 rod model due to the symmetry of the fuel rod and full model. From the evaluation of temperature for the 1/8 rod model at steady state, maximal error of 0.18 % was present relative to the ABAQUS result. Analysis for the transient problem using the fuel rod model resulted in the same as the variation of centerline temperature from the ABAQUS code during a hypothetical power transient. The distribution of heat flux for the entire cross section and surface was almost identical for the two codes

Fuel element reactivity worth in different rings of the IPR-R1 TRIGA reactor

Energy Technology Data Exchange (ETDEWEB)

Gomes do Prado Souza, Rose Mary

2008-10-29

The thermal power of the IPR-R1 TRIGA Reactor will be upgraded from 100 kW to 250 kW. Starting core: loaded with 59 aluminum cladded fuel elements; 1.34 $ excess reactivity; and 100 kW power. It is planned to go 2.5 times the power licensed, i.e., 250 kW. This forces to enlarge the reactivity level. Nuclear reactors must have sufficient excess reactivity to compensate the negative reactivity feedback effects caused by: the fuel temperature, fuel burnup, fission poisoning production, and to allow full power operation for predetermined period of time. To provide information for the calculation of the new core arrangement, the reactivity worth of some fuel elements in the core were measured as well as the determination of the core reactivity increase in the substitution of the original fuels, cladded with aluminium, for new ones, cladded with stainless steel. The reactivity worth of fuel element was measured from the difference in critical position of the control rods, calibrated by the positive period method, before and after the fuel element was withdrawn from the core. The magnitude of reactivity increase was determined when withdrawing the original Al-clad fuel (a little burned up) and the graphite elements, and inserting a fresh Al-clad fuel element, one by one. Experimental results indicated that to obtain enough reactivity excess to increase the rector power the addition of 4 new fuel elements in the core would be sufficient: - Substitution of 4 Al-clad fuel elements in ring C for fresh stainless steel clad fuel elements; - increase the reactivity {approx_equal} 4 x 6.5 = 26 cents; - The removed 4 Al-clad F. E. (a little burned up) put in the core periphery, ring F, replacing graphite elements; - add < 4 x 39 156 cents (39 cents was measured with a fresh F.E.). Neutron source was changed from position F7 to F8. Control and Safety rods were moved from ring D to C in order to increase their reactivity worth. Regulating rod was kept at the same position, F16. Four

Evaluation of spinal instrumentation rod bending characteristics for in-situ contouring.

Science.gov (United States)

Noshchenko, Andriy; Xianfeng, Yao; Armour, Grant Alan; Baldini, Todd; Patel, Vikas V; Ayers, Reed; Burger, Evalina

2011-07-01

Bending characteristics were studied in rods used for spinal instrumentation at in-situ contouring conditions. Five groups of five 6 mm diameter rods made from: cobalt alloy (VITALLIUM), titanium-aluminum-vanadium alloy (SDI™), β-titanium alloy (TNTZ), cold worked stainless steel (STIFF), and annealed stainless steel (MALLEABLE) were studied. The bending procedure was similar to that typically applied for in-situ contouring in the operating room and included two bending cycles: first--bending to 21-24° under load with further release of loading for 10 min, and second--bending to 34-37° at the previously bent site and release of load for 10 min. Applied load, bending stiffness, and springback effect were studied. Statistical evaluation included ANOVA, correlation and regression analysis. TNTZ and SDI™ rods showed the highest (p under load (p < 0.001). To reach the necessary bend angle after unloading, over bending should be 37-40% of the required angle in TNTZ and SDI™ rods, 27-30% in VITALLIUM and STIFF rods, and around 20% in MALLEABLE rods. Copyright © 2011 Wiley Periodicals, Inc.

Control-rod interference effects observed during reactor physics experiments with nuclear ship 'MUTSU'

International Nuclear Information System (INIS)

Itagaki, Masafumi; Miyoshi, Yoshinori; Gakuhari, Kazuhiko; Okada, Noboru; Sakai, Tomohiro.

1993-01-01

The control rods in the reactor of the nuclear ship MUTSU are classified into four groups: groups G1 and G2 are located in the central part of the core, while groups G3 and G4 are in the peripheral zone of the core. Several types of mutual interference effects among these control-rod groups were observed during reactor physics experiments with this reactor. During normal hot operations, positive shadowing was dominant between the G1 and G2 groups; the degree of the shadowing effect of one rod group depended on the position of the other rod group. Both positive and negative shadowing effects occurred between an inner rod group (G1 or G2) and an outer group (G3 or G4) depending on the three-dimensional arrangement of the control rods. The rod worths of G1 and G2 increased as a result of slight core burnup, about 1,400 MWd/t, mainly due to the decrease in shadowing effects resulting from a change in control-rod pattern. A three-dimensional diffusion calculation with internal control-rod boundary conditions has proved to be useful for analyzing these various interaction effects. (author)

The development and validation of control rod calculation methods

International Nuclear Information System (INIS)

Rowlands, J.L.; Sweet, D.W.; Franklin, B.M.

1979-01-01

Fission rate distributions have been measured in the zero power critical facility, ZEBRA, for a series of eight different arrays of boron carbide control rods. Diffusion theory calculations have been compared with these measurements. The normalised fission rates differ by up to about 30% in some regions, between the different arrays, and these differences are well predicted by the calculations. A development has been made to a method used to produce homogenised cross sections for lattice regions containing control rods. Calculations show that the method also reproduces the reaction rate within the rod and the fission rate dip at the surface of the rod in satisfactory agreement with the more accurate calculations which represent the fine structure of the rod. A comparison between diffusion theory and transport theory calculations of control rod reactivity worths in the CDFR shows that for the standard design method the finite mesh approximation and the difference between diffusion theory and transport theory (the transport correction) tend to cancel and result in corrections to be applied to the standard mesh diffusion theory calculations of about +- 2% or less. This result applies for mesh centred finite difference diffusion theory codes and for the arrays of natural boron carbide control rods for which the calculations were made. Improvements have also been made to the effective diffusion coefficients used in diffusion theory calculations for control rod followers and these give satisfactory agreement with transport theory calculations. (U.K.)

A study on performance of adjuster rod system and banking scheme in operational transient of CANDU-6 RUFIC core

International Nuclear Information System (INIS)

Kim, Soon Young; Suk, Ho Chun

2002-01-01

The performance of adjuster rod system in four operational transients of CANDU-6 RUFIC (Recovered Uranium Fuel In CANDU) core was preliminarily assessed, where the operational transients include startup after a short shutdown, startup after a poison-out shutdown, shim mode operation, and a stepback to 60% full power. The results of the preliminary assessment indicated that the adjuster rod system as currently designed and installed in the CANDU-6 NU (Natural Uranium) core will adequately meet the functional requirements in the RUFIC core. Comparing to the performance of adjuster rod system in the NU core, the total worth of the adjuster system in the RUFIC core is reduced, leading to less xenon override capability and shimming capability. In spite of the reduction of total worth, however, the overall performance of adjuster rod system in the operation transient of the RUFIC core is expected to still be satisfied. An alternative adjuster-banking scheme is also included in the assessment. The alternative adjuster-banking scheme involves rods in Bank 1 and Bank 7 being re-distributed within the two banks. The overall results from the transients studied indicated that the alternative banking scheme does show some better performance characteristics and merits

The treatment of absorber rod heterogeneity effects using homogeneous equivalent cross-sections and their application in large fast reactors

International Nuclear Information System (INIS)

Newton, T.D.

1988-01-01

This paper examines the application of homogeneous equivalent absorber rod cross-sections to the calculation of control rod anti-reactivities in large fast reactors. The method used to obtain the equivalent cross-sections is described and their validity in simple whole core geometry calculations is verified. Finally, they are employed in the calculation of control rod anti-reactivity worths in the Super Phenix 1 fast reactor and the results are compared with measured values. (author). 5 refs, 5 figs, 9 tabs

Optical coherence tomography for nondestructive evaluation of fuel rod degradation

International Nuclear Information System (INIS)

Renshaw, Jeremy B.; Jenkins, Thomas P.; Buckner, Benjamin D.; Friend, Brian

2015-01-01

Nuclear power plants regularly inspect fuel rods to ensure safe and reliable operation. Excessive corrosion can cause fuel failures which can have significant repercussions for the plant, including impacts on plant operation, worker exposure to radiation, and the plant's INPO rating. While plants typically inspect for fuel rod corrosion using eddy current techniques, these techniques have known issues with reliability in the presence of tenacious, ferromagnetic crud layers that can deposit during operation, and the nondestructive evaluation (NDE) inspection results can often be in error by a factor of 2 or 3. For this reason, alternative measurement techniques, such as Optical Coherence Tomography (OCT), have been evaluated that are not sensitive to the ferromagnetic nature of the crud. This paper demonstrates that OCT has significant potential to characterize the thickness of crud layers that can deposit on the surfaces of fuel rods during operation. Physical trials have been performed on simulated crud samples, and the resulting data show an apparent correlation between the crud layer thickness and the OCT signal

Optical coherence tomography for nondestructive evaluation of fuel rod degradation

Energy Technology Data Exchange (ETDEWEB)

Renshaw, Jeremy B., E-mail: jrenshaw@epri.com [Electric Power Research Institute, 1300 West WT Harris Blvd., Charlotte, NC 28262 (United States); Jenkins, Thomas P., E-mail: tjenkins@metrolaserinc.com; Buckner, Benjamin D., E-mail: tjenkins@metrolaserinc.com [MetroLaser, Inc., 22941 Mill Creek Drive, Laguna Hills, CA 92653 (United States); Friend, Brian [AREVA, Inc., 3315 Old Forest Road, Lynchburg, VA 24501 (United States)

2015-03-31

Nuclear power plants regularly inspect fuel rods to ensure safe and reliable operation. Excessive corrosion can cause fuel failures which can have significant repercussions for the plant, including impacts on plant operation, worker exposure to radiation, and the plant's INPO rating. While plants typically inspect for fuel rod corrosion using eddy current techniques, these techniques have known issues with reliability in the presence of tenacious, ferromagnetic crud layers that can deposit during operation, and the nondestructive evaluation (NDE) inspection results can often be in error by a factor of 2 or 3. For this reason, alternative measurement techniques, such as Optical Coherence Tomography (OCT), have been evaluated that are not sensitive to the ferromagnetic nature of the crud. This paper demonstrates that OCT has significant potential to characterize the thickness of crud layers that can deposit on the surfaces of fuel rods during operation. Physical trials have been performed on simulated crud samples, and the resulting data show an apparent correlation between the crud layer thickness and the OCT signal.

Reactivity anomalies in the FFTF [Fast Flux Test Facility

International Nuclear Information System (INIS)

Knutson, B.J.; Harris, R.A.

1987-04-01

Experience using an automated core reactivity monitoring technique at the Fast Flux Test Facility (FFTF) through eight operating cycles is described. This technique relies on comparing predicted to measured rod positions to detect any anomalous (or unpredicted) core reactivity changes. Reactivity worth predictions of core state changes (e.g., temperature and irradiation changes) and compensating control rod movements are required for the rod position comparison. A substantial data base now exists to evaluate changes in temperature reactivity feedback effects operational in the FFTF, rod worth changes due to core loading, temperature and irradiation effects and burnup effects associated with transmutation of fuel materials. This report summarizes preliminary work of correlating zero power and at-power rod worth measurement data, calculated burnup rates and rod worths using the latest ENDF/B-V cross section set for each cycle to evaluate the prediction models and attempt to resolve observed reactivity anomalies. 2 figs., 2 tabs

Report on the evaluation of the tritium producing burnable absorber rod lead test assembly. Revision 1

International Nuclear Information System (INIS)

1997-03-01

This report describes the design and fabrication requirements for a tritium-producing burnable absorber rod lead test assembly and evaluates the safety issues associated with tritium-producing burnable absorber rod irradiation on the operation of a commercial light water reactor. The report provides an evaluation of the tritium-producing burnable absorber rod design and concludes that irradiation can be performed within U.S. Nuclear Regulatory Commission regulations applicable to a commercial pressurized light water reactor

Evaluation of droplet deposition in rod bundle

International Nuclear Information System (INIS)

Ji, W.; Gu, C.Y.; Anglart, H.

1997-01-01

Deposition model for droplets in gas droplet two-phase flow in rod bundle is developed in this work using the Lagrangian method. The model is evaluated in a 9-rod bundle geometry. The deposition coefficient in the bundle geometry are compared with that in round tube. The influences of the droplet size and gas mass flow rate on deposition coefficient are investigated. Furthermore, the droplet motion is studied in more detail by dividing the bundle channel into sub-channels. The results show that the overall deposition coefficient in the bundle geometry is close to that in the round tube with the diameter equal to the bundle hydraulic diameter. The calculated deposition coefficient is found to be higher for higher gas mass flux and smaller droplets. The study in the sub-channels show that the ratio between the local deposition coefficient for a sub-channel and the averaged value for the whole bundle is close to a constant value, deviations from the mean value for all the calculated cases being within the range of ±13%. (author)

Eliminating the human element and the drudgery from control-rod calibrations

Energy Technology Data Exchange (ETDEWEB)

Ruby, L; Wang, H -K [Univ. of California, Berkeley (United States)

1974-07-01

The Berkeley TRIGA Mark III Reactor has three distinct reflector arrangements, depending on the position of the core in the pool. The control rods must be calibrated in each position, making 12 rod calibrations required, in all. To eliminate the human element and the drudgery involved in this repetitious task, a computer-assisted semi-automatic method has been devised to perform the necessary period methods, and to produce the resultant rod-calibration curves. The method is based on the use of a signal from the linear-power-channel recorder to feed a voltage comparator which generates a pulse at a preselected voltage 'B' and also at '1.50V'. The 2 pulses are used to start and stop pulses for an electronic timer, which easily measures the time difference to 0.01 second. The comparator actually consists of two such pulse-pair generating circuits, so that 2 measurements of t{sub 50} can be obtained on each range of the linear-power channel. Before the comparator is used for a series of rod calibrations, the voltage discrimination levels are checked with a precision voltage source to verify that they are set at 3.50, 5.25, 6.00, and 9.00 volts. Corrections in the discrimination levels can be made by means of front-panel potentiometer adjustments. As voltage is gradually increased past each of the pre-set discrimination levels, a panel light comes on, indicating that a pulse has been formed. The comparator circuit also accepts a reset command from a push button held in the hand of the reactor operator, which command is then converted into an electrical reset signal for the electronic timer. The system provides non-prejudiced measurements for t{sub 50} as short as 5 seconds, with no concern about pen lag. The only manipulation of the data is to determine the best value of t{sub 50}, which is done by averaging those values which agree to within 0.1 second. The program ''RODCALN'' is used to calculate the rod worth remaining (in dollar units) versus control rod position

Preliminary scoping study of some neutronic aspects of new shim safety rods for a typical 5 MW research reactor by Monte Carlo simulation

Energy Technology Data Exchange (ETDEWEB)

Shoushtari, M.K.; Kakavand, T. [Faculty of Science, University of Zanjan, P.O. BOX 1415, Zanjan (Iran, Islamic Republic of); Ghaforian, H. [Faculty of Science and Technology of Marine, P.O. BOX 212 Tehran (Iran, Islamic Republic of); Kiai, S.M. Sadat [Nuclear Science and Technology Research Institute (NSTR), Nuclear Science Research, A.E.O.I. P.O. BOX 14155-1339, Tehran (Iran, Islamic Republic of)], E-mail: sadatkiai@yahoo.com

2009-02-15

A Monte Carlo simulation of a typical 5 MW research reactor (TRR) was carried out using MCNP4C code. The geometry of the reactor core was modeled including the details of all fuel elements, control rods, all irradiation channels, graphite reflectors, reactor pool and thermal column. The model predicted neutron flux distributions within the core, control rod (CR) worth, core reactivity ({rho}), shutdown margin, and some kinetic parameters when the control rod insert or withdraw. This study was carried out to reduce blockage probability of shim safety rod (SSR)s of the TRR. Two introduced more blackness SSRs were chosen and made thinner in a way adequate blackness, in comparison to the present rods, achieved.

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

Evaluation of rod insertion issue for NPP Krsko

International Nuclear Information System (INIS)

Gunstek, A.; Kurincic, B.

1998-01-01

The last couple of years incident with control rods sticking in lower part of the fuel assemblies have been reported of several reactor operators and fuel vendors throughout of the world. Several activities were initiated immediately to determine the root cause of incomplete rod insertion. The purpose of this activities were to collect plants trip history data and testing results, review of available worldwide experience, review of plant operation and fuel management, detailed review of manufacturing and material property and to maintain detailed mechanical model. In this paper, we will present activities in Nuclear Power Plant Krsko which have been performed after NRC initiated the Root Cause Process (NRC Bulletin 96-01). NPP Krsko has not experienced rod insertion anomaly yet but anyway the additional tests were carried out. Rod drop time measurements that were performed normally at beginning of cycle at nominal temperature and pressure (HSB mode) have been extended also to end of cycle. Rod drop time, velocity of dropped rods and magnitudes of the initial recoil bounces vs. burnup were also analyzed. Also RCCA drag test with upper internals in place and drive shafts attached to RCCAs has been performed since then. At last two outages (1997 and 1998) drag test were carried out with digital scale meter to gather additional information. In addition to that, the reload core design has been performed with new constrains on rodded fuel assembly burnup as proposed by the industry.(author)

Control rod homogenization in heterogeneous sodium-cooled fast reactors

International Nuclear Information System (INIS)

Andersson, Mikael

2016-01-01

The sodium-cooled fast reactor is one of the candidates for a sustainable nuclear reactor system. In particular, the French ASTRID project employs an axially heterogeneous design, proposed in the so-called CFV (low sodium effect) core, to enhance the inherent safety features of the reactor. This thesis focuses on the accurate modeling of the control rods, through the homogenization method. The control rods in a sodium-cooled fast reactor are used for reactivity compensation during the cycle, power shaping, and to shutdown the reactor. In previous control rod homogenization procedures, only a radial description of the geometry was implemented, hence the axially heterogeneous features of the CFV core could not be taken into account. This thesis investigates the different axial variations the control rod experiences in a CFV core, to determine the impact that these axial environments have on the control rod modeling. The methodology used in this work is based on previous homogenization procedures, the so-called equivalence procedure. The procedure was newly implemented in the PARIS code system in order to be able to use 3D geometries, and thereby be take axial effects into account. The thesis is divided into three parts. The first part investigates the impact of different neutron spectra on the homogeneous control-rod cross sections. The second part investigates the cases where the traditional radial control-rod homogenization procedure is no longer applicable in the CFV core, which was found to be 5-10 cm away from any material interface. In the third part, based on the results from the second part, a 3D model of the control rod is used to calculate homogenized control-rod cross sections. In a full core model, a study is made to investigate the impact these axial effects have on control rod-related core parameters, such as the control rod worth, the capture rates in the control rod, and the power in the adjacent fuel assemblies. All results were compared to a Monte

Feasibility evaluation of x-ray imaging for measurement of fuel rod bowing in CFTL test bundles

International Nuclear Information System (INIS)

Baker, S.P.

1980-06-01

The Core Flow Test Loop (CFTL) is a high temperature, high pressure, out-of-reactor helium-circulating system. It is designed for detailed study of the thermomechanical performance, at prototypic steady-state and transient operating conditions, of electrically heated rods that simulate segments of core assemblies in the Gas-Cooled Fast Breeder reactor demonstration plant. Results are presented of a feasibility evaluation of x-ray imaging for making measurements of the displacement (bowing) of fuel rods in CFTL test bundles containing electrically heated rods. A mock-up of a representative CFTL test section consisting of a test bundle and associated piping was fabricated to assist in this evaluation

The use of eddy current testing for nuclear fuel rods cladding evaluation

International Nuclear Information System (INIS)

Silva Junior, Silverio F. da; Alencar, Donizete A.; Brito, Mucio Jose D. de

2007-01-01

Nuclear fuel rods cladding must be tested after their manufacture and during their operational life. This paper describes a study about the use of eddy current test method as a nondestructive tool for nuclear fuel rods cladding evaluation. The experiments were carried out using two different probes: an external probe and an internal probe. The main goal was to verify the sensitivity of the eddy current test system, to develop calibration and reference standards and to establish the main capabilities and limitations presented by this test method for this application. (author)

Methods for reactor physics calculations for control rods in fast reactors

International Nuclear Information System (INIS)

Grimstone, M.J.; Rowlands, J.L.

1988-12-01

The IAEA Specialists' Meeting on ''Methods for Reactor Physics Calculations for Control Rods in Fast Reactors'' was held in Winfrith, United Kingdom, on 6-8 December, 1988. The meeting was attended by 23 participants from nine countries. The purpose of the meeting was to review the current calculational methods and their accuracy as assessed by theoretical studies and comparisons with measurements, and then to identify the requirements for improved methods or additional studies and comparisons. The control rod properties or effects to be considered were their reactivity worths, their effect on the power distribution through the core, and the reaction rates and energy deposition both within and adjacent to the rods. The meeting was divided into five sessions, in the first of which each national delegation presented a brief overview of their programme of work on calculational methods for fast reactor control rods. In the next three sessions a total of seventeen papers were presented describing calculational methods and assessments of their accuracy. The final session was a discussion to draw conclusions regarding the current status of methods and the further developments and validation work required. A separate abstract was prepared for each of the 23 papers presented at the meeting. Refs, figs and tabs

Radioactive lightning rods: radiologic evaluation and regulatory policy related to its use in Cuba

International Nuclear Information System (INIS)

Lopez Forteza, Yamil; Quevedo Garcia, Jose R.; Diaz Guerra, Pedro I.; Cruz Dumenico, Gonzalez; Fuente Puch, Andres de la

2001-01-01

The radioactive lightning rod employment for the protection of facilities against atmospheric discharges reached its maximum splendor in the eighties. It was in fact at the end of this decade when the technical considerations related to the justification of this practice finally conclude that the production of such teams was abolished. For the regulatory authorities, however, it continues having validity the question related to the control of lightning rod still in use as well as the question related to the establishment of a coherent with the international practice national policy. The paper shows the results of the last 10 years of control of the radioactive lightning rod use in Cuba and the radiological evaluation carried out on the base of this experience. Lastly, it exposes the regulatory policy referred to the employment of the radioactive lightning rod in the country. (author)

Control rod shadowing and anti-shadowing effects in a large gas-cooled fast reactor

International Nuclear Information System (INIS)

Girardin, G.; Chawla, R.; Rimpault, G.; Coddington, P.

2007-01-01

An investigation of control rod shadowing and anti-shadowing (interaction) effects has been carried out in the context of a design study of the control rod pattern for the large 2400 MWth Generation IV Gas-cooled Fast Reactor (GFR). For the calculations, the deterministic code system ERANOS-2.0 has been used, in association with a full core model including a European Fast Reactor (EFR)-type pattern for the control rods. More specifically, the core contains a total of 33 control (CSD) and safety (DSD) rods implemented in three banks: -1) a first bank of 6 CSD rods, placed at 64 cm from core centre in the inner fuel zone (Pu content 16.3 % vol.), -2) a safety bank consisting of 9 DSD rods, at an average distance of 118 cm, and -3) a third bank with 18 CSD rods, placed at 171 cm, i.e. at the interface between the inner and outer (Pu content 19.2 % vol.) core regions. Each control rod has been modelled as a homogeneous material containing 90%-enriched B 4 C, steel and helium. Considerable shadowing effects have been observed between the first bank and the safety bank, as also between individual rods within the first bank. Large anti-shadowing effects take place in an even greater number of the studied rod configurations. The largest interaction is between the two CSD banks, the anti-shadowing value being 46% in this case, implying that the total rod worth is increased by a factor of almost 2 when compared to the sum of the individual bank values. Additional investigations have been performed, in particular the computation of the first order eigenvalue and the eigenvalue separation. The main finding is that the interactions are lower when one of the control rod banks is located at a radial position corresponding to half the core radius. (authors)

Comparison of Wims-Aecl / Dragon / RFSP and MCNP results with Zed-2 measurements for control device worth and reactor kinetics - 037

International Nuclear Information System (INIS)

Pencer, J.; Choy Wong, F.; Bromley, B.P.; Atfield, J.; Zeller, M.

2010-01-01

This paper summarizes comparisons between MCNP5 and WIMS-AECL / DRAGON / RFSP calculations and experimental results obtained from the Zero Energy Deuterium (ZED-2) critical facility at AECL Chalk River Laboratories. MCNP5 and WIMS-AECL / DRAGON / RFSP were used to calculate reactivity worths for two reactivity devices, a mechanical zone controller (MZC) and shut-off rod (SOR) in a lattice similar to that of the ACR-1000 R . WIMS-AECL / DRAGON / RFSP was also used to obtain kinetics parameters for a transient based on a rod drop of a ZED-2 standby absorber rod (SAR). ZED-2 experiments were performed using 43-element ACR Low Enriched Uranium (ACR-LEU) fuel bundles with H 2 O- or air-cooled fuel bundles arranged in a 24-cm pitch square lattice. Calculations with MCNP5 gave biases in device worths that were within 0.2 mk of measured values, while WIMS-AECL / DRAGON / RFSP gave values that were within 0.3 mk of measured values. Transient analyses using the CERBERUS module within RFSP yielded a total delayed neutron fraction (β) that was within 4% of the value derived by point kinetics analysis of experimental data. The corresponding delayed photo-neutron fraction (β photo-neutron ) from CERBERUS was within 5% of that derived by point kinetics. This study has helped quantify the agreement between calculation and measurement for codes that are used in the safety analysis of the ACR-1000 reactor. Results demonstrate good agreement in code predictions. (authors)

Reactivity determination of the Al2O3-B4C burnable poison as a function of its concentration in the IPEN/MB-01 reactor

International Nuclear Information System (INIS)

Giada, Marino Reis

2005-01-01

Burnable poison rods made of Al 2 O 3 -B 4 C pellets with different concentrations of 10 B have been manufactured for a set of experiments in the IPEN/MB-01 zero-power reactor. The experiments evaluated the reactivity of the burnable poison rods as a function of the 10 B concentration, and the shadowing effect on the control rod reactivity worth as a function of the distance between the burnable position rods and the control rod. The results showed that the burnable poison rods have a non-linear behavior as function of the 10 B concentration, starting to reach an asymptotic value for concentrations higher than 7 g/cm 3 of 10 B. The shadowing effect on the control rods was substantial. When the burnable poison rods were beside the control rod, its reactivity worth decreased as much as 30 %, and when they were 10,5 cm distant, the control rod worth decreased by 7 %. The MCNP results for the burnable poison reactivity effects agreed within experimental errors with the measured values. (author)

Benchmark Evaluation of HTR-PROTEUS Pebble Bed Experimental Program

International Nuclear Information System (INIS)

Bess, John D.; Montierth, Leland; Köberl, Oliver

2014-01-01

Benchmark models were developed to evaluate 11 critical core configurations of the HTR-PROTEUS pebble bed experimental program. Various additional reactor physics measurements were performed as part of this program; currently only a total of 37 absorber rod worth measurements have been evaluated as acceptable benchmark experiments for Cores 4, 9, and 10. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the 235 U enrichment of the fuel, impurities in the moderator pebbles, and the density and impurity content of the radial reflector. Calculations of k eff with MCNP5 and ENDF/B-VII.0 neutron nuclear data are greater than the benchmark values but within 1% and also within the 3σ uncertainty, except for Core 4, which is the only randomly packed pebble configuration. Repeated calculations of k eff with MCNP6.1 and ENDF/B-VII.1 are lower than the benchmark values and within 1% (~3σ) except for Cores 5 and 9, which calculate lower than the benchmark eigenvalues within 4σ. The primary difference between the two nuclear data libraries is the adjustment of the absorption cross section of graphite. Simulations of the absorber rod worth measurements are within 3σ of the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments

Analysis of rod drop and pulsed source measurements of reactivity in the Winfrith SGHWR

International Nuclear Information System (INIS)

Brittain, I.

1970-05-01

Reactivity measurements by the rod-drop and pulsed source methods in the Winfrith SGHWR are seriously affected by spatial harmonics. A method of calculation is described which enables the spatial harmonics to be calculated in non-uniform cores in two or three dimensions, and thus allows a much more rigorous analysis of the experimental results than the usual point model. The method is used to analyse all the rod-drop measurements made during commissioning of the Winfrith SGHWR, and to comment on the results of pulsed source measurements. The reactivity worths of banks of ten and twelve shut-down tubes deduced from rod-drop and pulsed source experiments are in satisfactory agreement with each other and also with AIMAZ calculated values. The ability to calculate higher spatial harmonics in nonuniform cores is thought to be new, and may have a wider application to reactor kinetics through the method of Modal Analysis. (author)

Evaluation of control rod motion simulator research reactors

International Nuclear Information System (INIS)

Sanda

2010-01-01

Motion simulator has been carried out testing of the reactor control rod using a servomotor. Reactor control rod motion at any point should be in the right position, one of the motors that can move in a precise and correct the servo motor. To ensure that the servo motor to move in accordance with the desired program, then the servomotor function test for motor work to ensure the performance of the appliance. Tests carried out on meshes stress disorder, the load is stable within a certain period and travel time safety control rod up and down, travel time regulating control rods up and down and travel time compensation control rods up and down. In testing the breakdown voltage Vout nets at 24 V, 6.5 A with 12 Ω load deviation obtained V0 = V1 = 0.1% and 0.65% and for the stability of the load in a certain time deviation V = 0.7125%, next to the breakdown voltage Vout nets at 12V, 4.2 A with a 6 Ω load deviation obtained V0 = V1 = 0.275% and 1.158% for the stability of the load in a certain time deviation V = 1.463% and the net-voltage noise nets on Vout 24 V, 4.5 A with 12 Ω load deviation obtained V0 = V1 = 0.196% and 0.496% and for the stability of the load in a certain time deviation V = 0.3625%. While the travel time of a safety control rod up and down, up and down the regulator and compensation rise and fall showed a steady linear graph. The results show that the performance of the servo motor is very stable with the working area below the tolerance limit, it is 5% - 10%. (author)

An empirical evaluation of two theoretically-based hypotheses on the directional association between self-worth and hope.

Science.gov (United States)

McDavid, Lindley; McDonough, Meghan H; Smith, Alan L

2015-06-01

Fostering self-worth and hope are important goals of positive youth development (PYD) efforts, yet intervention design is complicated by contrasting theoretical hypotheses regarding the directional association between these constructs. Therefore, within a longitudinal design we tested: (1) that self-worth predicts changes in hope (self theory; Harter, 1999), and (2) that hope predicts changes in self-worth (hope theory; Snyder, 2002) over time. Youth (N = 321; Mage = 10.33 years) in a physical activity-based PYD program completed surveys 37-45 days prior to and on the second day and third-to-last day of the program. A latent variable panel model that included autoregressive and cross-lagged paths indicated that self-worth was a significant predictor of change in hope, but hope did not predict change in self-worth. Therefore, the directional association between self-worth and hope is better explained by self-theory and PYD programs should aim to enhance perceptions of self-worth to build perceptions of hope. Copyright © 2015 The Foundation for Professionals in Services for Adolescents. Published by Elsevier Ltd. All rights reserved.

Construction and initial validation of the self-worth protection scale.

Science.gov (United States)

Thompson, Ted; Dinnel, Dale L

2003-03-01

The self-worth theory of achievement motivation holds that in certain circumstances students stand to gain by deliberately withdrawing effort. When failure occurs despite effort, students are likely to conclude that failure resulted from lack of ability. Thus, withdrawing effort offers a defence against conclusions of low ability, thereby protecting self-worth. We undertook to assess the psychometric properties of the Self-Worth Protection Scale (SWPS). Data were obtained from 243 participants (Study 1) and 411 participants (Study 2) enrolled in undergraduate psychology courses at a university in the United States. We administered a number of scales, including the SWPS and scales assessing a fear of negative evaluation, academic self-esteem, uncertain global self-evaluations, self-handicapping, and causal uncertainty. Exploratory factor analysis indicated a three-factor solution (ability doubts, the importance of ability as a criterion of self-worth, and an avoidance orientation) utilising 33 of the original 44 items. A confirmatory factor analysis indicated that this three-factor solution was a poor fit of the data. After modifying the model, a confirmatory factor analysis indicated that a three-factor solution with 26 of the original items and a higher order factor of self-worth protection was an adequate fit of the data. Reliability measures were acceptable for the three subscales and total score. The total score of the SWPS was positively correlated with theoretically related constructs, demonstrating construct validity. The SWPS appears to be a psychometrically sound scale to assist in identifying individuals who manifest self-worth protection in achievement situations.

FARST: A computer code for the evaluation of FBR fuel rod behavior under steady-state/transient conditions

International Nuclear Information System (INIS)

Nakamura, M.; Sakagami, M.

1984-01-01

FARST, a computer code for the evaluation of fuel rod thermal and mechanical behavior under steady-state/transient conditions has been developed. The code characteristics are summarized as follows: (I) FARST evaluates the fuel rod behavior under the transient conditions. The code analyzes thermal and mechanical phenomena within a fuel rod, taking into account the temperature change in coolant surrounding the fuel rod. (II) Permanent strains such as plastic, creep and swelling strains as well as thermoelastic deformations can be analyzed by using the strain increment method. (III) Axial force and contact pressure which act on the fuel stack and cladding are analyzed based on the stick/slip conditions. (IV) FARST used a pellet swelling model which depends on the contact pressure between pellet and cladding, and an empirical pellet relocation model, designated as 'jump relocation model'. The code was successfully applied to analyses of the fuel rod irradiation data from pulse reactor for nuclear safety research in Cadarache (CABRI) and pulse reactor for nuclear safety research in Japan Atomic Energy Research Institute (NSRR). The code was further applied to stress analysis of a 1000 MW class large FBR plant fuel rod during transient conditions. The steady-state model which was used so far gave the conservative results for cladding stress during overpower transient, but underestimated the results for cladding stress during a rapid temperature decrease of coolant. (orig.)

Comparable Worth, Job Evaluation and Wage Discrimination: The Employer Approaches Wage Gap Issues in the 1980's.

Science.gov (United States)

Leach, Daniel E.

1984-01-01

The role of women in the work force and the wages paid to women workers have become major employment discrimination issues of the 1980's. Comparable worth, wage discrimination, and the existence and possible influence of sex-related factors in wage administration systems, which include formalized job evaluation schemes, are discussed. (MLW)

Determination of the most reactivity control rod by pseudo-harmonics perturbation method; Determinacao da barra de controle mais reativa usando o metodo de pseudo-harmonicos

Energy Technology Data Exchange (ETDEWEB)

Freire, Fernando S.; Silva, Fernando C.; Martinez, Aquilino S. [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear]. E-mail: ffreire@con.ufrj.br; fernando@con.ufrj.br; aquilino@.con.ufrj.br

2005-07-01

Frequently it is necessary to compute the change in core multiplication caused by a change in the core temperature or composition. Even when this perturbation is localized, such as a control rod inserted into the core, one does not have to repeat the original criticality calculation, but instead we can use the well-known pseudo-harmonics perturbation method to express the corresponding change in the multiplication factor in terms of the neutron flux expanded in the basis vectors characterizing the unperturbed core. Therefore we may compute the control rod worth to find the most reactivity control rod to calculate the fast shutdown margin. In this thesis we propose a simple and precise method to identify the most reactivity control rod. (author)

Monte Carlo simulation of a research reactor with nominal power of 7 MW to design new control safety rods

Energy Technology Data Exchange (ETDEWEB)

Shoushtari, M.K.; Kakavand, T. [Faculty of Science, University of Zanjan, Zanjan, P.O. Box 451-313 (Iran, Islamic Republic of); Sadat Kiai, S.M., E-mail: sadatkiai@yahoo.co [Nuclear Science and Technology Research Institute (NSTR), Nuclear Science Research, A.E.O.I., P.O. Box 14155-1339, Tehran (Iran, Islamic Republic of); Ghaforian, H. [Faculty of Science and Technology of Marine, Tehran (Iran, Islamic Republic of)

2010-03-01

The Monte Carlo simulation has been established for a research reactor with nominal power of 7 MW. A detailed model of the reactor core was employed including standard and control fuel elements, reflectors, irradiation channels, control rods, reactor pool and thermal column. The following physical parameters of reactor core were calculated for the present LEU core: core reactivity (rho), control rod (CR) worth, thermal and epithermal neutron flux distributions, shutdown margin and delayed neutron fraction. Reduction of unfavorable effects of blockage probability of control safety rod (CSR)s in their interiors because of not enough space in their sites, and lack of suitable capabilities to fabricate very thin plates for CSR cladding, is the main aim of the present study. Making the absorber rod thinner and CSR cladding thicker by introducing a better blackness absorbing material and a new stainless steel alloy, respectively, are two studied ways to reduce the effects of mentioned problems.

Application of hafnium hydride control rod to large sodium cooled fast breeder reactor

Energy Technology Data Exchange (ETDEWEB)

Ikeda, Kazumi, E-mail: kazumi_ikeda@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Moriwaki, Hiroyuki, E-mail: hiroyuki_moriwaki@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Ohkubo, Yoshiyuki, E-mail: yoshiyuki_okubo@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Iwasaki, Tomohiko, E-mail: tomohiko.iwasaki@qse.tohoku.ac.jp [Department of Quantum Science and Energy Engineering, Tohoku University, Aoba, Aramaki, Aoba-ku, Sendai-shi, Miyagi-ken 980-8579 (Japan); Konashi, Kenji, E-mail: konashi@imr.tohoku.ac.jp [Institute for Materials Research, Tohoku University, Narita-cho, Oarai-machi, Higashi-Ibaraki-gun, Ibaraki-ken 311-1313 (Japan)

2014-10-15

Highlights: • Application of hafnium hydride control rod to large sodium cooled fast breeder reactor. • This paper treats application of an innovative hafnium hydride control rod to a large sodium cooled fast breeder reactor. • Hydrogen absorption triples the reactivity worth by neutron spectrum shift at H/Hf ratio of 1.3. • Lifetime of the control rod quadruples because produced daughters of hafnium isotopes are absorbers. • Nuclear and thermal hydraulic characteristics of the reactor are as good as or better than B-10 enriched boron carbide. - Abstract: This study treats the feasibility of long-lived hafnium hydride control rod in a large sodium-cooled fast breeder reactor by nuclear and thermal analyses. According to the nuclear calculations, it is found that hydrogen absorption of hafnium triples the reactivity by the neutron spectrum shift at the H/Hf ratio of 1.3, and a hafnium transmutation mechanism that produced daughters are absorbers quadruples the lifetime due to a low incineration rate of absorbing nuclides under irradiation. That is to say, the control rod can function well for a long time because an irradiation of 2400 EFPD reduces the reactivity by only 4%. The calculation also reveals that the hafnium hydride control rod can apply to the reactor in that nuclear and thermal characteristics become as good as or better than 80% B-10 enriched boron carbide. For example, the maximum linear heat rate becomes 3% lower. Owing to the better power distribution, the required flow rate decreases approximately by 1%. Consequently, it is concluded on desk analyses that the long lived hafnium hydride control rod is feasible in the large sodium-cooled fast breeder reactor.

ISOS: A job evaluation system to implement comparable worth

Directory of Open Access Journals (Sweden)

Albert Corominas

2008-01-01

Full Text Available A fair pay structure is an essential element of the personnel policy of a firm. If the pay structure is perceived as arbitrary by the members of the staff, it becomes a cause of disturbance of the labor relations. Particularly, a pay structure is unfair if it discriminates against women. Job evaluation is a traditional tool used by companies to assist in the process of determining pay structures that can be also useful to detect and combat wage discrimination, since allow determining whether two jobs are of comparable worth or not. Although there are many kinds of systems, authors agree when defining point factor methods as the most appropriate and fair job evaluation systems. However, even being well defined from a technical point of view, most existing systems give discriminatory results regarding to gender. ISOS, a new job evaluation system which is described in this paper, has been designed, with the aim to define a neutral system with regard to gender, based on present jobs characteristics, existing job evaluation systems and job description questionnaires, international experts’ knowledge and a wide body of literature on gender discrimination and its relation with job evaluation. Using ISOS can contribute to detect, combat and eliminate part of the existing wage discrimination in general and, in particular, against women. ISOS includes all aspects of the work so no characteristics are omitted. The system can be applied in any company and to evaluate any job, and offers flexibility to be adapted to the specific characteristics of an organization. ISOS can also be used to detect and combat wage discrimination. Furthermore, characteristics of present jobs, such as cross-training or flexible working time, are also included so the system can be considered innovative in a very traditional field of industrial engineering.

Performance Evaluation of the New Fork-Absorbers of RSG-GAS Control Rod

International Nuclear Information System (INIS)

Slamet Wiranto; Purwadi; Arif Hidayat; Agus Sanjaya

2012-01-01

During the operation of RSG-GAS reactor, it has been replaced 8 fork-absorber by the new absorber from PT. Batan Teknologi. After almost 5 years under utilization it is important to be evaluated to determine the physical condition and its performance, which is still in good condition and functioning according to the requirements of its operations. The evaluation has been carried out by studying and analyzing the data of the fork-absorber utilization in the the reactor core. The fork absorber data consist of visual inspection, control rod drop time measurement and control rod reactivity and safety margin measurement for each operation cycle. Through the observation up to date with the operating cycle of 79, could be concluded that the fork-absorber condition is still good, and has ability, to support the operation until ± 660 MWD/cycle, which is characterized by obtaining the value of ρ-excess is sufficient for operation, with a large safety margin. (author)

Neutronic analysis of absorbing materials for the control rod system in reactor ALLEGRO

Energy Technology Data Exchange (ETDEWEB)

Cajko, Frantisek; Secansky, Michal; Chrebet, Tomas; Zajac, Radoslav; Darilek, Petr [VUJE, a.s., Trnava (Slovakia)

2016-09-15

Experimental reactor ALLEGRO is a gas cooled fast reactor in the design stage. The current design of its reactivity control system is based on control rods filled with boron carbide as the absorber. Because of disadvantages connected to high boron enrichment a possibility of using other absorbent materials was explored to lower the boron enrichment and increase the worth of the control rods. The results of neutronic Monte-Carlo analyses in a computational supercell are presented in this paper. Three absorbent materials most suitable for a use in reactor ALLEGRO (B{sub 4}C, EuB{sub 6} and ReB{sub 2}) have been analysed also in a full core model. A possible benefit of a neutron trap concept is explored as well but materials with satisfactory neutronic properties proved to be not suitable for expected high temperatures in the reactor.

Preliminary Content Evaluation of the North Anna High Burn-Up Sister Fuel Rod Segments for Transportation in the 10-160B and NAC-LWT

Energy Technology Data Exchange (ETDEWEB)

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

2016-08-09

The U.S. Department of Energy’s (DOE’s) Used Fuel Disposition Campaign (UFDC) Program has transported high-burnup nuclear sister fuel rods from a commercial nuclear power plant for purposes of evaluation and testing. The evaluation and testing of high-burnup used nuclear fuel is integral to DOE initiatives to collect information useful in determining the integrity of fuel cladding for future safe transportation of the fuel, and for determining the effects of aging, on the integrity of UNF subjected to extended storage and subsequent transportation. The UFDC Program, in collaboration with the U.S. Nuclear Regulatory Commission and the commercial nuclear industry, has obtained individual used nuclear fuel rods for testing. The rods have been received at Oak Ridge National Laboratory (ORNL) for both separate effects testing (SET) and small-scale testing (SST). To meet the research objectives, testing on multiple 6 inch fuel rod pins cut from the rods at ORNL will be performed at Pacific Northwest National Laboratory (PNNL). Up to 10 rod equivalents will be shipped. Options were evaluated for multiple shipments using the 10-160B (based on 4.5 rod equivalents) and a single shipment using the NAC-LWT. Based on the original INL/Virginia Power transfer agreement, the rods are assumed to 152 inches in length with a 0.374-inch diameter. This report provides a preliminary content evaluation for use of the 10-160B and NAC-LWT for transporting those fuel rod pins from ORNL to PNNL. This report documents the acceptability of using these packagings to transport the fuel segments from ORNL to PNNL based on the following evaluations: enrichment, A2 evaluation, Pu-239 FGE evaluation, heat load, shielding (both gamma and neutron), and content weight/structural evaluation.

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-07-01

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

Control rod studies in small and medium sized fast reactors

International Nuclear Information System (INIS)

John, T.M.; Mohanakrishnan, P.; Mahalakshmi, B.; Singh, R.S.

1988-01-01

Control rods are the primary safety mechanism in the operation of fast reactors. Neutronic parameters associated with the control rods have to be evaluated precisely for studying the behaviour of the reactor under various operating conditions. Control rods are strong neutron absorbers discretely distributed in the reactor core. Accurate estimation of control rod parameters demand, in principle transport theory solutions in exact geometry. But computer codes for such evaluations usually consume exorbitantly large computer time and memory for even a single parameter evaluation. During the design of reactors, evaluation of these parameters will be required for many configurations of control rods. In this paper, the method used at Indira Gandhi Centre for Atomic Research for estimating the parameters associated with control rods is presented. Diffusion theory solutions were used for computations. A scheme using three dimensional geometry represented by triangular meshes and diffusion theory solutions in few energy groups for control rod parameter evaluation is presented. This scheme was employed in estimating the control rod parameters in a 500 Mw(e) fast reactor. Error due to group collapsing is estimated by comparing with 25 group calculations in three dimensions for typical cases. (author). 5 refs, 4 figs, 3 tabs

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.)

A new physics design of control safety rods for prototype fast breeder reactor

International Nuclear Information System (INIS)

Devan, K.; Riyas, A.; Alagan, M.; Mohanakrishnan, P.

2008-01-01

The absorber rods of 500 MWe prototype fast breeder reactor (PFBR), which is under construction at Kalpakkam, have been designed to provide sufficient shutdown margin during normal and accidental conditions for ensuring the safe shut down. There are nine control and safety rods (CSR) and 3 diverse safety rods (DSR). Absorber material used is initially 65% enriched B 4 C. Based on the reported experiments in PHENIX reactor and design of absorber rods in SUPERPHENIX, the design of CSR is modified by introducing 20 cm length natural B 4 C at the top and bottom of absorber column and maintaining the remaining portion with 65% enriched B 4 C. This design ensures sufficient shutdown margin (SDM) during normal operation and also during the one stuck rod condition. For comparison of the above two designs, a CSR of 57% of enrichment was considered which gives the same worth as the revised CSR design with natural B 4 C sections in top and bottom. There is significant savings in the initial inventory of enriched B 4 C for CSR. The annual requirement of enriched boron also reduces. This new CSR can last for about 5 cycles, based on its clad life. But, it is planned to be replaced after every 3 cycles (1 cycle equals 180 efpd) of operation due to radiation damage effects in hexcan D9 steel. Use of ferritic steel for hexcan can extend the life of CSR to 5 cycles

Performance evaluation of the Loviisa advanced type fuel rods

International Nuclear Information System (INIS)

Ranta-Puska, K.; Pihlatie, M.

2001-01-01

The fuel vendor TVEL has supplied to Loviisa WWER-440 power plant six lead assemblies of an advanced type which have profiling of the fuel enrichment, demountability of the assembly and a reduced shroud wall thickness. The pool side examination programme of these assemblies is underway including visual inspections, diameter and length measurements between operation cycles, and end-of-life fission gas release measurements, determined from 85 Kr activity in the plenum. Complementary evaluations and testing of models are done with the ENIGMA fuel performance code. The diameters of the corner rods have decreased to 30 μm during the first cycle and 40 to 70 μm after two cycles (with rod burnups of 24-30 MWd/kgU). The extent of creep-down is generally as expected, and agrees with the creep model adjusted for Russian Zr1%Nb cladding type and the Loviisa coolant and neutron flux conditions. The gap closure and reversed hoop strain are to be awaited during the third cycle so the new data will be an interesting validation exercise for the model and ENIGMA. Calculated temperatures stay low, and therefore low fission gas release fractions are anticipated as well

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.)

Design and Operational Evaluation of the Traffic Management Advisor at the Ft. Worth Air Route Traffic Control Center

Science.gov (United States)

Swenson, Harry N.; Vincent, Danny; Tobias, Leonard (Technical Monitor)

1997-01-01

NASA and the FAA have designed and developed and an automation tool known as the Traffic Management Advisor (TMA). The system was operationally evaluated at the Ft. Worth Air Route Traffic Control Center (ARTCC). The TMA is a time-based strategic planning tool that provides Traffic Management Coordinators and En Route Air Traffic Controllers the ability to efficiently optimize the capacity of a demand impacted airport. The TMA consists of trajectory prediction, constraint-based runway scheduling, traffic flow visualization and controllers advisories. The TMA was used and operationally evaluated for forty-one rush traffic periods during a one month period in the Summer of 1996. The evaluations included all shifts of air traffic operations as well as periods of inclement weather. Performance data was collected for engineering and human factor analysis and compared with similar operations without the TMA. The engineering data indicates that the operations with the TMA show a one to two minute per aircraft delay reduction during rush periods. The human factor data indicate a perceived reduction in en route controller workload as well as an increase in job satisfaction. Upon completion of the evaluation, the TMA has become part of the normal operations at the Ft. Worth ARTCC.

Self-reported "worth it" rating of aesthetic surgery in social media.

Science.gov (United States)

Domanski, Mark C; Cavale, Naveen

2012-12-01

A wide variety of surveys have been used to validate the satisfaction of patients who underwent aesthetic surgery. However, such studies are often limited by patient number and number of surgeons. Social media now allows patients, on a large scale, to discuss and rate their satisfaction with procedures. The views of aesthetic procedures patients expressed in social media provide unique insight into patient satisfaction. The "worth it" percentage, average cost, and number of respondents were recorded on October 16, 2011, for all topics evaluated on the aesthetic procedure social media site www.realself.com . Procedures were divided into categories: surgical, liposuction, nonsurgical, and dental. For each group, procedures with the most respondents were chosen and ordered by "worth it" score. A literature search was performed for the most commonly rated surgical procedures and the satisfaction rates were compared. A total of 16,949 evaluations of 159 aesthetic surgery topics were recorded. A correlation between cost of the procedure and percentage of respondents indicating that the procedure was "worth it" was not found. The highest-rated surgical procedure was abdominoplasty, with 93 % of the 1,589 self-selected respondents expressing that abdominoplasty was "worth it." The average self-reported cost was $8,400. The highest-rated nonsurgical product was Latisse, with 85 % of 231 respondents reporting it was "worth it" for an average cost of $200. The satisfaction scores in the literature for commonly rated surgical procedures ranged from 62 to 97.6 %. No statistically significant correlations between literature satisfaction scores and realself.com "worth it" scores were found. Abdominoplasty had the highest "worth it" rating among aesthetic surgical procedures. Aesthetic surgeons should be wary that satisfaction scores reported in the literature might not correlate with commonly achieved results. Social media has opened a new door into how procedures are

Evaluation of the contamination risk by 241AM from lightning rods disposed at uncontrolled garbage dump

International Nuclear Information System (INIS)

Marumo, Julio Takehiro

2006-01-01

Radioactive lightning rods were manufactured in Brazil until 1989, when the licenses for using radioactive sources in these products were lifted by the national nuclear authority. Since then, radioactive devices have been replaced by Franklin type one and collected as radioactive waste. However, only 23 percent of the estimated total number of installed rods was delivered to Brazilian Nuclear Commission (Comissao Nacional de Energia Nuclear - CNEN). This situation is of concern as there is a possibility of the rods being discarded as domestic waste, considering that in Brazil, 63.6 percent of the municipal solid waste is disposed at uncontrolled garbage dump, according to Instituto Brasileiro de Geografia e Estatistica (IBGE) in 2000. In addition, americium, the most common employed radionuclide, is classified as a high toxicity element, when ingested or inhaled. In the present study, it was performed migration experiments of Am-241 by lysimeter system in order to evaluate the risk of contamination caused by radioactive lightning rods disposed as a common solid waste. Sources removed from lightning rods were placed inside lysimeters filled with organic waste, collected at the restaurant of Instituto de Pesquisas Energeticas e Nucleares, IPEN-CNEN/SP, and the generated leachate was periodically analyzed to determine its characteristics such as pH, redox potential, solid content and concentration of the radioactive material. Microbial growth was also evaluated by counting the number of colony forming units. The equivalent dose to members of the public has been calculated considering the ingestion of drinking water, the most probable mode of exposure. The final result was about 145 times below the effective dose limit of 1 mSv.year-1 for members of the public, established by the International Commission on Radiological Protection (ICRP), demonstrating that the risk caused by lightning rods disposed at uncontrolled garbage dump is low. (author)

An outline of the methods which will be used to evaluate shutdown margins, control rod worth and local criticality problems for the HTR - power reactors

Energy Technology Data Exchange (ETDEWEB)

Dworak, A

1971-09-28

The methods to calculate control rod efficiency outlined in this paper were the results of an investigation carried out by A. Dworak and C. Hunt on the Dragon Charge I Core 1. There the theoretical results agreed very good with the experimental measurements. This paper gives a brief outline of the methods in question and summarizes the calculations done up to now.

Puerto Rican kindergartners' self-worth as coded from the Attachment Story Completion Task: correlated with other self-evaluation measures and ratings of child behavior toward mothers and peers.

Science.gov (United States)

Gullón-Rivera, Ángel L

2013-01-01

This multi-method multi-informant study assessed 105 Puerto Rican kindergartners' sense of self-worth in family relationships as coded from their responses to the Attachment Story Completion Task (ASCT). The ASCT scores were compared with responses to two other age-appropriate self-evaluation measures (the Cassidy Puppet Interview and the Pictorial Scales of Social Acceptance). Correlations of children's scores on the three self-measures with maternal ratings of the mother-child relationship and teacher ratings of the child's prosocial behavior with peers were then compared. ASCT self-worth and Puppet Interview scores were strongly correlated with each other and both were modestly related to the pictorial social acceptance scales. All three measures were significantly associated with maternal and teacher reports of child behavior, but the strongest correlations were obtained with the ASCT. Coding the ASCT in terms of self-worth appears to be a promising approach for evaluating young children's (vicariously expressed) self-worth in family relationships.

Calculation of homogenized Pickering NGS stainless steel adjuster rod neutron cross-sections using conservation of reaction rates

Energy Technology Data Exchange (ETDEWEB)

Robinson, R C [Atlantic Nuclear Services Ltd. (Canada); Tran, F [Ontario Hydro, Pickering, ON (Canada). Pickering Generating Station

1996-12-31

A homogenization methodology for calculation of reactivity device incremental cross-sections has been developed using reaction rate conservation (RRC). A heterogeneous transport calculation of flux was utilised to produce the homogenized cross-sections for a finite difference two group diffusion code. The RRC cross-sections have been shown to improve significantly the prediction of reactivity worth for stainless steel adjuster rods installed in Pickering NGS reactors. (author). 10 refs., 3 tabs., 6 figs.

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.

LMR design concepts for transuranic management in low sodium void worth cores

International Nuclear Information System (INIS)

Hill, R.N.

1991-01-01

The fuel cycle processing techniques and hard neuron spectrum of the Integral Fast Reactor (IFR) metal fuel cycle have favorable characteristics for the management of transuranics; and the wide range of breeding characteristics available in metal fuelled cores provides for flexibility in transuranic management strategy. Previous studies indicate that most design options which decrease the breeding ratio also show a decrease in sodium void worth; therefore, low void worths are achievable in transuranic burning (low breeding ratio) core designs. This paper describes numerous trade studies assessing various design options for a low void worth transuranic burner core. A flat annular core design appears to be a promising concept; the high leakage geometry yields a low breeding ratio and small sodium void worth. To allow flexibility in breeding characteristics, alternate design options which achieve fissile self-sufficiency are also evaluated. A self-sufficient core design which is interchangeable with the burner core and maintains a low sodium void worth is developed. 13 refs., 1 fig., 4 tabs

LMR design concepts for transuranic management in low sodium void worth cores

International Nuclear Information System (INIS)

Hill, R.N.

1991-01-01

The fuel cycle processing techniques and hard neutron spectrum of the integral Fast Reactor (IFR) metal fuel cycle have favorable characteristics for the management of transuranics; and the wide range of breeding characteristics available in metal fuelled cores provides for flexibility in transuranic management strategy. Previous studies indicate that most design options which decrease the breeding ratio also allow a decrease in sodium void worth; therefore, low void worths are achievable in transuranic burning (low breeding ratio) core designs. This paper describes numerous trade studies assessing various design options for a low void worth transuranic burner core. A flat annular core design appears to be a promising concept; the high leakage geometry yields a low breeding ratio and small sodium void worth. To allow flexibility in breeding characteristics, alternate design options which achieve fissile self-sufficiency are also evaluated. A self-sufficient core design which is interchangeable with the burner core and maintains a low sodium void worth is developed. (author)

Evaluation of the effect of probe design parameters on ECT signal and development of eddy current probe for irradiated fuel rods

International Nuclear Information System (INIS)

Kwank, S. W.; Han, Y. K.; Woo, S. K.; Kim, T. W.; Park, J. Y.; Kim, B. J.; Park, J. Y.

1999-01-01

Eddy current test(ECT) is used to inspect not only the failed fuel rods but also peripheral rods during repairing of the failed fuel rods, to detect internal defects in irradiated fuel rods which could not be detected by ultrasonic test and visual test, and to obtain the data for determining the root cause of fuel rod failure. This study evaluates the effect of properties of test article, irradiated fuel rods, on the impedance diagram in order to reduce the difficulty of ECT signal analysis. The optimum eddy current probe design conditions for inspecting the irradiated fuel rods, is estimate by using experimental equations and the probe is manufactured based on the estimated conditions. The performance of developed eddy current probe and the optimum conditions is proved through characteristic comparison experiment with the probe purchased from the foreign vendor

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.)

International symposium on fuel rod simulators: development and application

Energy Technology Data Exchange (ETDEWEB)

McCulloch, R.W. (comp.)

1981-05-01

Separate abstracts are included for each of the papers presented concerning fuel rod simulator operation and performance; simulator design and evaluation; clad heated fuel rod simulators and fuel rod simulators for cladding investigations; fuel rod simulator components and inspection; and simulator analytical modeling. Ten papers have previously been input to the Energy Data Base.

Significant others and contingencies of self-worth: activation and consequences of relationship-specific contingencies of self-worth.

Science.gov (United States)

Horberg, E J; Chen, Serena

2010-01-01

Three studies tested the activation and consequences of contingencies of self-worth associated with specific significant others, that is, relationship-specific contingencies of self-worth. The results showed that activating the mental representation of a significant other with whom one strongly desires closeness led participants to stake their self-esteem in domains in which the significant other wanted them to excel. This was shown in terms of self-reported contingencies of self-worth (Study 1), in terms of self-worth after receiving feedback on a successful or unsatisfactory performance in a relationship-specific contingency domain (Study 2), and in terms of feelings of reduced self-worth after thinking about a failure in a relationship-specific contingency domain (Study 3). Across studies, a variety of contingency domains were examined. Furthermore, Study 3 showed that failing in an activated relationship-specific contingency domain had negative implications for current feelings of closeness and acceptance in the significant-other relationship. Overall, the findings suggest that people's contingencies of self-worth depend on the social situation and that performance in relationship-specific contingency domains can influence people's perceptions of their relationships.

Analysis of control rod behavior based on numerical simulation

International Nuclear Information System (INIS)

Ha, D. G.; Park, J. K.; Park, N. G.; Suh, J. M.; Jeon, K. L.

2010-01-01

The main function of a control rod is to control core reactivity change during operation associated with changes in power, coolant temperature, and dissolved boron concentration by the insertion and withdrawal of control rods from the fuel assemblies. In a scram, the control rod assemblies are released from the CRDMs (Control Rod Drive Mechanisms) and, due to gravity, drop rapidly into the fuel assemblies. The control rod insertion time during a scram must be within the time limits established by the overall core safety analysis. To assure the control rod operational functions, the guide thimbles shall not obstruct the insertion and withdrawal of the control rods or cause any damage to the fuel assembly. When fuel assembly bow occurs, it can affect both the operating performance and the core safety. In this study, the drag forces of the control rod are estimated by a numerical simulation to evaluate the guide tube bow effect on control rod withdrawal. The contact condition effects are also considered. A full scale 3D model is developed for the evaluation, and ANSYS - commercial numerical analysis code - is used for this numerical simulation. (authors)

EXPERIMENTAL EVALUATION OF THE FULLY LOADED ELK RIVER REACTOR

Energy Technology Data Exchange (ETDEWEB)

Fisher, J. R.; Diaz, A.

1963-06-15

The loading and testing program of the Elk River Reactor confirmed the predicted values. The measured cold, clean excess reactivity agrees to 2% and the control rod worths to 1% of the calculated values. The reactivity for various core loadings and rod positions is tabulated. The effects of spiked elements on the reactivity and radial peak-toaverage power ratio were studied. (D.L.C.)

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

Development of cutting device for irradiated fuel rod

International Nuclear Information System (INIS)

Lee, E. P.; Jun, Y. B.; Hong, K. P.; Min, D. K.; Lee, H. K.; Su, H. S.; Kim, K. S.; Kwon, H. M.; Joo, Y. S.; Yoo, K. S.; Joo, J. S.; Kim, E. K.

2004-01-01

Post Irradiation Examination(PIE) on irradiated fuel rods is essential for the evaluation of integrity and irradiation performance of fuel rods of commercial reactor fuel. For PIE, fuel rods should be cut very precisely. The cutting positions selected from NDT data are very important for further destructive examination and analysis. A fuel rod cutting device was developed witch can cut fuel rods longitudinal very precisely and can also cut the fuels into the same length rod cuts repeatedly. It is also easy to remove the fuel cutting powder after cutting works and it can extend the life time of cutting device and lower the contamination level of hot cell

Uncertainty of Doppler reactivity worth due to uncertainties of JENDL-3.2 resonance parameters

Energy Technology Data Exchange (ETDEWEB)

Zukeran, Atsushi [Hitachi Ltd., Hitachi, Ibaraki (Japan). Power and Industrial System R and D Div.; Hanaki, Hiroshi; Nakagawa, Tuneo; Shibata, Keiichi; Ishikawa, Makoto

1998-03-01

Analytical formula of Resonance Self-shielding Factor (f-factor) is derived from the resonance integral (J-function) based on NR approximation and the analytical expression for Doppler reactivity worth ({rho}) is also obtained by using the result. Uncertainties of the f-factor and Doppler reactivity worth are evaluated on the basis of sensitivity coefficients to the resonance parameters. The uncertainty of the Doppler reactivity worth at 487{sup 0}K is about 4 % for the PNC Large Fast Breeder Reactor. (author)

LOFT advanced fuel rod instrumentation development

International Nuclear Information System (INIS)

Billeter, T.R.; Brown, R.L.; Chan, A.I.Y.; Day, C.K.; Meyers, S.C.; Sheen, E.M.; Stringer, J.L.

1978-01-01

Advanced fuel rod instrumentation for the Loss of Fluid Test (LOFT) reactor is being developed by the Hanford Engineering Development Laboratory for the Nuclear Regulatory Commission. This effort calls for development of sensors to measure fuel rod axial motion, fuel centerline temperature (to 2200 0 C), fuel rod plenum gas pressure (to 2500 psig), and plenum gas temperature (to 1500 0 F). A parallel test and evaluation of several modified commercial sensors was undertaken and will result in commercial availability of the final qualified sensors. Necessary test facilities were prepared for the development and evaluation effort. Tests to date indicate a three coil Linear Variable Differential Transformer (LVDT), operated from temperature compensating signal source and processing electronics, will meet the desired requirements

Semi-empirical evaluation studies on PCMI for the Fugen fuel rod

International Nuclear Information System (INIS)

Domoto, Kazushige; Kaneko, Mitsunobu; Takeuchi, Kiyoshi.

1980-03-01

Fugen, 165 MWe prototype of a heavy water moderated boiling water cooled reactor, has been well operated since March 1979. In order to establish PCIOMR for Fugen fuels semi-empirical evaluation code to analyze PCMI during power transient of the fuel rod has been developed. In this paper, followings are described 1) general scope of the development work 2) description of the modelling 3) some results of analysis on out pile and in pile tests. (author)

An evaluation of control rod motion simulator of research reactor

International Nuclear Information System (INIS)

Sanda

2010-01-01

Motion simulator for rod control research reactor has been carried out using a servo motor. Reactor rod motion control at any point should be in the right position, one of the motors that can move in a precise and correct is the servo motor. To ensure that the servo motor to move in accordance with the desired program, then the servo motor function test should be carried out to ensure having good performance. Tests carried out on meshes stress disorder, the load is stable within a certain period and travel time safety control rod up and down, travel time regulating control rods up and down and travel time compensation control rods up and down. In testing the breakdown voltage V out nets at 24 V, 6.5 A with 12 Q load deviation obtained V0= V1 = 0.1% and 0.65% and for the stability of the load in a certain time deviation V = 0.7125% , next to the breakdown voltage V out nets at 12 V, 4.2 A with a 6 Q load deviation obtained V0= V1 = 0.275% and 1.158% for the stability of the load in a certain time deviation V = 1.463% and the net-voltage noise nets on V out 24 V, 4.5 A with 12 Q load deviation obtained V0 = V1 = 0.196% and 0.496% and for the stability of the load in a certain time deviation V = 0.3625%. While the travel time of a safety control rod up and down, up and down the regulator and compensation rise and fall showed a steady linear graph. The results show that the performance of the servo motor is very stable with the working area below the tolerance limit, it is 5% - 10%.(author)

Preliminary nuclear design for test MOX Fuel rods

Energy Technology Data Exchange (ETDEWEB)

Joo, Hyung Kook; Kim, Taek Kyum; Jeong, Hyung Guk; Noh, Jae Man; Cho, Jin Young; Kim, Young Il; Kim, Young Jin; Sohn, Dong Seong

1997-10-01

As a part of activity for future fuel development project, test MOX fuel rods are going to be loaded and irradiated in Halden reactor core as a KAERI`s joint international program with Paul Scherrer Institute (PSI). PSI will fabricate test MOX rods with attrition mill device which was developed by KAERI. The test fuel assembly rig contains three MOX rods and three inert matrix rods. One of three MOX rods will be fabricated by BNFL, the other two MOX fuel rods will be manufacturing jointly by KAERI and PSI. Three inert matrix fuel rods will be fabricated with Zr-Y-Er-Pu oxide. Neutronic evaluation was preliminarily performed for test fuel assembly suggested by PSI. The power distribution of test fuel rod in test fuel assembly was analyzed for various fuel rods position in assembly and the depletion characteristic curve for test fuel was also determined. The fuel rods position in test fuel assembly does not effect the rod power distribution, and the proposal for test fuel rods suggested by PSI is proved to be feasible. (author). 2 refs., 13 tabs., 16 figs.

Design and evaluation of an on-line fuel rod assay device for an HTGR fuel refabrication plant

International Nuclear Information System (INIS)

Rushton, J.E.; Allen, E.J.; Chiles, M.M.; Jenkins, J.D.

1979-11-01

Refabricated HTGR fuel rods will contain from approx. 0.15 to 0.5 g 233 U and/or 235 U. The fuel rods are approx. 16 mm in diameter and 62 mm long. A typical commercial fuel refabrication facility will have six fuel rod production lines, each producing approximately one fuel rod every 4 seconds at design capacity. One on-line assay device will be present for each two production lines. The relative standard deviation in an individual fuel rod fissile material measurement must be less than 3% to satisfy process and quality control requirements. Systematic errors must be kept less than approx. 0.3% for fissile material measured in fuel rods produced over two months to satisfy material accountability requirements. Several nondestructive assay (NDA) methods were investigated. Because the gamma-ray activity of the refabricated fuel is relatively high due to the presence of 232 U in the fuel and because the gamma-ray activity is not directly related to total or fissile uranium content, NDA methods employing gamma-ray detection did not appear practicable. A method using thermal neutron irradiation and fast-fission neutron detection was selected. An experimental assay device was fabricated based on this NDA method. Experiments were performed to determine the precision and accuracy of the measurements and to investigate potential interferences and systematic errors. Operating procedures were evaluated, and analysis procedures were identified

Development of an ATCA IPMI Controller Mezzanine Board and its usage on an ATCA ROD Evaluator board for the ATLAS LAr upgrade

CERN Document Server

"DUMONT DAYOT, N; The ATLAS collaboration; "CAP, S; "DUMONT DAYOT, N; "FOURNIER, L; "LETENDRE, N; "PERROT, G; "Wingerter, I

2011-01-01

In the context of the LHC upgrades, a new Read-Out Driver (ROD) board for the ATLAS LAr calorimeter is being developed. xTCA (Advanced/Micro Telecom Computing Architecture) is becoming a standard in high energy physics and is a serious candidate for future readout systems. We will present our current developments to master ATCA and to integrate a large number of very high speed links (96 links/8.5 Gbps) on a ROD Evaluator ATCA board. To manage our ROD Evaluator, we have developed a versatile ATCA IPMI controller for ATCA boards which is FPGA Mezzanine Card (FMC) compliant.

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.)

A Preliminary Assessment of the Adjuster Rod Depletion Effect in the CANDU Reactor

International Nuclear Information System (INIS)

Kim, Yonghee; Roh, Gyuhong; Kim, Won Young; Kim, Hak Sung; Park, Joo Hwan

2008-01-01

Lifetime of the Wolsong-1 CANDU reactor, which will be shutdown in April, 2009. Major reactor components such as the pressure tube are to be replaced and it is expected that the CANDU reactor can be operated for additional 25-30 years. Meanwhile, all the reactivity devices including the adjuster rods (ADJ) are supposed to be continuously used without any change. In the CANDU reactor, 21 stainless steel (SS) ADJs are used to control the core power distribution and compensate for some reactivity loss during several abnormal cases. The ADJs are normally fully inserted and the SS absorber should undergo a slow depletion through neutron irradiation for a long time. In April, 2009, the accumulated FPY (Full Power Day) of Wolsong-1 is about 23 years. Depletion of ADJs should result in a smaller ADJ worth and a higher fuel burnup and the core power distribution should also be affected by the ADJ depletion. In this work, the effects of the ADJ depletion have been assessed in terms of ADJ worth, time-average core characteristics

Attachment styles and contingencies of self-worth.

Science.gov (United States)

Park, Lora E; Crocker, Jennifer; Mickelson, Kristin D

2004-10-01

Previous research on attachment theory has focused on mean differences in level of self-esteem among people with different attachment styles. The present study examines the associations between attachment styles and different bases of self-esteem, or contingencies of self-worth, among a sample of 795 college students. Results showed that attachment security was related to basing self-worth on family support. Both the preoccupied attachment style and fearful attachment style were related to basing self-worth on physical attractiveness. The dismissing attachment style was related to basing self-worth less on others' approval, family support, and God's love.

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.)

Reactivity balance for a soluble boron-free small modular reactor

Directory of Open Access Journals (Sweden)

Lezani van der Merwe

2018-06-01

Full Text Available Elimination of soluble boron from reactor design eliminates boron-induced reactivity accidents and leads to a more negative moderator temperature coefficient. However, a large negative moderator temperature coefficient can lead to large reactivity feedback that could allow the reactor to return to power when it cools down from hot full power to cold zero power. In soluble boron-free small modular reactor (SMR design, only control rods are available to control such rapid core transient.The purpose of this study is to investigate whether an SMR would have enough control rod worth to compensate for large reactivity feedback. The investigation begins with classification of reactivity and completes an analysis of the reactivity balance in each reactor state for the SMR model.The control rod worth requirement obtained from the reactivity balance is a minimum control rod worth to maintain the reactor critical during the whole cycle. The minimum available rod worth must be larger than the control rod worth requirement to manipulate the reactor safely in each reactor state. It is found that the SMR does have enough control rod worth available during rapid transient to maintain the SMR at subcritical below k-effectives of 0.99 for both hot zero power and cold zero power. Keywords: Control Rod Worth, Reactivity Balance, Reactivity Feedback, Small Modular Reactor, Soluble Boron Free

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.)

Neutronic evaluation of annular fuel rods to assemblies 13 x 13, 14 x 14 and 15 x 15

International Nuclear Information System (INIS)

Silva, Raphael H.M.; Ramos, Mario C.; Velasquez, Carlos E.; Silva, Clarysson A.M. da; Pereira, Cláubia; Costa, Antonella L.

2017-01-01

Research and development in nuclear reactor field has been proposed a new concept of fuel rod such as annular shape. The design of the annular fuel rods allows the coolant flow through the inner and outer side of it. Such project was proposed as an alternative to the traditional fuel rods used in LWR reactors. This new geometry allows an increase in power density in the reactor core with greater heat transfer from the fuel to the coolant which reduces the temperature in central region of the rod, in which a better configuration and dimension of fuel elements are aimed due to improvement of cooling in possible replacement of PWR traditional rods for annular rods. The aim of this work is to evaluate the neutronic parameters of fuel element with annular fuel rods where three configurations were studied: 13 x 13, 14 x 14 and 15 x 15. The goal is compare the neutronic between the advanced and the standard fuel assembly 16 x 16. In these studies, the external dimension and the moderator to fuel volume ratio (V M /V F ) of standard 16 x 16 is the same in all annular fuels assemblies. The MCNPX 2.6.0 (Monte Carlo N-Particle eXtended – version 2.6.0) code was used in all simulations. After all procedures, the annular fuel assemblies 13 have obtained greater neutronics parameters and were selected to more neutronics simulations. (author)

Neutronic evaluation of annular fuel rods to assemblies 13 x 13, 14 x 14 and 15 x 15

Energy Technology Data Exchange (ETDEWEB)

Silva, Raphael H.M.; Ramos, Mario C.; Velasquez, Carlos E.; Silva, Clarysson A.M. da; Pereira, Cláubia; Costa, Antonella L., E-mail: rapha.galo@hotmail.com, E-mail: marc5663@gmail.com, E-mail: carlosvelcab@hotmail.com, E-mail: clarysson@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: antonella@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-07-01

Research and development in nuclear reactor field has been proposed a new concept of fuel rod such as annular shape. The design of the annular fuel rods allows the coolant flow through the inner and outer side of it. Such project was proposed as an alternative to the traditional fuel rods used in LWR reactors. This new geometry allows an increase in power density in the reactor core with greater heat transfer from the fuel to the coolant which reduces the temperature in central region of the rod, in which a better configuration and dimension of fuel elements are aimed due to improvement of cooling in possible replacement of PWR traditional rods for annular rods. The aim of this work is to evaluate the neutronic parameters of fuel element with annular fuel rods where three configurations were studied: 13 x 13, 14 x 14 and 15 x 15. The goal is compare the neutronic between the advanced and the standard fuel assembly 16 x 16. In these studies, the external dimension and the moderator to fuel volume ratio (V{sub M}/V{sub F}) of standard 16 x 16 is the same in all annular fuels assemblies. The MCNPX 2.6.0 (Monte Carlo N-Particle eXtended – version 2.6.0) code was used in all simulations. After all procedures, the annular fuel assemblies 13 have obtained greater neutronics parameters and were selected to more neutronics simulations. (author)

Fuel rod pressure in nuclear power reactors: Statistical evaluation of the fuel rod internal pressure in LWRs with application to lift-off probability

Energy Technology Data Exchange (ETDEWEB)

Jelinek, Tomas

2001-02-01

In this thesis, a methodology for quantifying the risk of exceeding the Lift-off limit in nuclear light water power reactors is outlined. Due to fission gas release, the pressure in the gap between the fuel pellets and the cladding increases with burnup of the fuel. An increase in the fuel-clad gap due to clad creep would be expected to result in positive feedback, in the form of higher fuel temperatures, leading to more fission gas release, higher rod pressure, etc, until the cladding breaks. An increase in the fuel-clad gap that leads to this positive feedback is a phenomenon called Lift-off and is a limitation that must be considered in the fuel core management. Lift-off is a consequence of very high internal fuel rod pressure. The internal fuel rod pressure is therefore used as a Lift-off indicator. The internal fuel rod pressure is closely connected to the fission gas release into the fuel rod plenum and is thus used to increase the database. It is concluded that the dominating error source in the prediction of the pressure in Boiling Water Reactors (BWR), is the power history. There is a bias in the fuel pressure prediction that is dependent on the fuel rod position in the fuel assembly for BWRs. A methodology to quantify the risk of the fuel rod internal pressure exceeding a certain limit is developed; the risk is dependent of the pressure prediction and the fuel rod position. The methodology is based on statistical treatment of the discrepancies between predicted and measured fuel rod internal pressures. Finally, a methodology to estimate the Lift-off probability of the whole core is outlined.

Refabricated and instrumented fuel rods

International Nuclear Information System (INIS)

Silberstein, K.

2005-01-01

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

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

Critical mass, rod values and reactivity coefficients for Rapsodie; Masse critique, valeur des barres et coefficients de reactivite de rapsodie

Energy Technology Data Exchange (ETDEWEB)

Stevens, L; Gourdon, J [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

1967-07-01

Besides a brief general description, the report contains a description and discussion of the aims, the methods used and the results of critical mass, rod worth and static reactivity coefficient measurements on the Rapsodie reactor. (authors) [French] Apres une breve description generale, le rapport decrit et discute le but, les methodes employees et les resultats des mesures de masse critique, de reactivite des barres et des coefficients de reactivite statiques du reacteur RAPSODIE. (auteurs)

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.

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

Evaluation of local free carrier concentrations in individual heavily-doped GaN:Si micro-rods by micro-Raman spectroscopy

Science.gov (United States)

Mohajerani, M. S.; Khachadorian, S.; Schimpke, T.; Nenstiel, C.; Hartmann, J.; Ledig, J.; Avramescu, A.; Strassburg, M.; Hoffmann, A.; Waag, A.

2016-02-01

Three-dimensional III-nitride micro-structures are being developed as a promising candidate for the future opto-electrical devices. In this study, we demonstrate a quick and straight-forward method to locally evaluate free-carrier concentrations and a crystalline quality in individual GaN:Si micro-rods. By employing micro-Raman mapping and analyzing lower frequency branch of A1(LO)- and E1(LO)-phonon-plasmon-coupled modes (LPP-), the free carrier concentrations are determined in axial and planar configurations, respectively. Due to a gradual doping profile along the micro-rods, a highly spatially resolved mapping on the sidewall is exploited to reconstruct free carrier concentration profile along the GaN:Si micro-rods. Despite remarkably high free carrier concentrations above 1 × 1020 cm-3, the micro-rods reveal an excellent crystalline quality, without a doping-induced stress.

Ultrasonic Nondestructive Evaluation of Pultruded Rod Stitched Efficient Unitized Structure (PRSEUS) During Large-Scale Load Testing and Rod Push-Out Testing

Science.gov (United States)

Johnston, Patrick H.; Juarez, Peter D.

2016-01-01

The Pultruded Rod Stitched Efficient Unitized Structure (PRSEUS) is a structural concept developed by the Boeing Company to address the complex structural design aspects associated with a pressurized hybrid wing body (HWB) aircraft configuration. The HWB has long been a focus of NASA's environmentally responsible aviation (ERA) project, following a building block approach to structures development, culminating with the testing of a nearly full-scale multi-bay box (MBB), representing a segment of the pressurized, non-circular fuselage portion of the HWB. PRSEUS is an integral structural concept wherein skins, frames, stringers and tear straps made of variable number of layers of dry warp-knit carbon-fiber stacks are stitched together, then resin-infused and cured in an out-of-autoclave process. The PRSEUS concept has the potential for reducing the weight and cost and increasing the structural efficiency of transport aircraft structures. A key feature of PRSEUS is the damage-arresting nature of the stitches, which enables the use of fail-safe design principles. During the load testing of the MBB, ultrasonic nondestructive evaluation (NDE) was used to monitor several sites of intentional barely-visible impact damage (BVID) as well as to survey the areas surrounding the failure cracks after final loading to catastrophic failure. The damage-arresting ability of PRSEUS was confirmed by the results of NDE. In parallel with the large-scale structural testing of the MBB, mechanical tests were conducted of the PRSEUS rod-to-overwrap bonds, as measured by pushing the rod axially from a short length of stringer.

Failed fuel rod detection method by ultrasonic wave

International Nuclear Information System (INIS)

Takamatsu, Masatoshi; Muraoka, Shoichi; Ono, Yukio; Yasojima, Yujiro.

1990-01-01

Ultrasonic wave signals sent from an ultrasonic receiving element are supplied to an evaluation circuit by way of a gate. A table for gate opening and closing timings at the detecting position in each of the fuel rods in a fuel assembly is stored in a memory. A fuel rod is placed between an ultrasonic transmitting element and the receiving element to determine the positions of the transmitting element and the receiving element by positional sensors. The opening and closing timings at the positions corresponding to the result of the detection are read out from the table, and the gates are opened and closed by the timing. This can introduce the ultrasonic wave signals transmitted through a control rod always to the evaluation circuit passing through the gate. Accordingly, the state of failure of the fuel rod can be detected accurately. (I.N.)

Verification study of thorium cross section in MVP calculation of thorium based fuel core using experimental data

International Nuclear Information System (INIS)

Mai, V. T.; Fujii, T.; Wada, K.; Kitada, T.; Takaki, N.; Yamaguchi, A.; Watanabe, H.; Unesaki, H.

2012-01-01

Considering the importance of thorium data and concerning about the accuracy of Th-232 cross section library, a series of experiments of thorium critical core carried out at KUCA facility of Kyoto Univ. Research Reactor Inst. have been analyzed. The core was composed of pure thorium plates and 93% enriched uranium plates, solid polyethylene moderator with hydro to U-235 ratio of 140 and Th-232 to U-235 ratio of 15.2. Calculations of the effective multiplication factor, control rod worth, reactivity worth of Th plates have been conducted by MVP code using JENDL-4.0 library [1]. At the experiment site, after achieving the critical state with 51 fuel rods inserted inside the reactor, the measurements of the reactivity worth of control rod and thorium sample are carried out. By comparing with the experimental data, the calculation overestimates the effective multiplication factor about 0.90%. Reactivity worth of the control rods evaluation using MVP is acceptable with the maximum discrepancy about the statistical error of the measured data. The calculated results agree to the measurement ones within the difference range of 3.1% for the reactivity worth of one Th plate. From this investigation, further experiments and research on Th-232 cross section library need to be conducted to provide more reliable data for thorium based fuel core design and safety calculation. (authors)

Multi-indicator Evaluation System for Broadsword, Rod, Sword and Spear Athletes Based on Analytic Hierarchy Process

Science.gov (United States)

Luo, Lin

2017-08-01

In the practical selection of Wushu athletes, the objective evaluation of the level of athletes lacks sufficient technical indicators and often relies on the coach’s subjective judgments. It is difficult to accurately and objectively reflect the overall quality of the athletes without a fully quantified indicator system, thus affecting the level improvement of Wushu competition. The analytic hierarchy process (AHP) is a systemic analysis method combining quantitative and qualitative analysis. This paper realizes structured, hierarchized and quantified decision-making process of evaluating broadsword, rod, sword and spear athletes in the AHP. Combing characteristics of the athletes, analysis is carried out from three aspects, i.e., the athlete’s body shape, physical function and sports quality and 18 specific evaluation indicators established, and then combining expert advice and practical experience, pairwise comparison matrix is determined, and then the weight of the indicators and comprehensive evaluation coefficient are obtained to establish the evaluation model for the athletes, thus providing a scientific theoretical basis for the selection of Wushu athletes. The evaluation model proposed in this paper has realized the evaluation system of broadsword, rod, sword and spear athletes, which has effectively improved the scientific level of Wushu athletes selection in practical application.

Reliabilty worth: Development of a relationship with outage magnitude, duration and frequency

International Nuclear Information System (INIS)

Turner, F.P.P.; Katrichak, A.M.; Dwyer, A.; Edwards, D.; Ibrahim, A.

1994-01-01

British Columbia Hydro's Worth Project Team was founded to determine values for reliability for reference in evaluation of investment and operating decisions. Work to date has produced key preliminary values for specific outages and concepts for the shape of the relationship between value and these determinates of reliability worth, frequency, magnitude and duration. These values and concepts are described. The values are developed through an iterative, trial and refinement approach. The approach incorporates direct input from customers, common sense and judgement, and micro- and macro-economic concepts. Reliability worth values for reduced or prevented outages are presented for residential, commercial, small industrial and mixed sectors and various outage durations. Reliability worth values were obtained through customer surveys. Limitations of the reliability worth value are numerous and are listed. Study of cost vs magnitude of interruption using microeconomic models has shown that costly system improvements to reduce the possibility of widespread outages may not be justified. The case of exceptionally large area outages (blackouts) is examined. The cost vs frequency relationship was examined in terms of the economic concept of utility or satisfaction. Different loss/frequency characteristics are demonstrated for different customer classes. Customer value for reduced outage duration is expressed in a curve with flatter slope than that for eliminated outages. 2 refs., 6 figs

Local heterogeneity effects on small-sample worths

International Nuclear Information System (INIS)

Schaefer, R.W.

1986-01-01

One of the parameters usually measured in a fast reactor critical assembly is the reactivity associated with inserting a small sample of a material into the core (sample worth). Local heterogeneities introduced by the worth measurement techniques can have a significant effect on the sample worth. Unfortunately, the capability is lacking to model some of the heterogeneity effects associated with the experimental technique traditionally used at ANL (the radial tube technique). It has been suggested that these effects could account for a large portion of what remains of the longstanding central worth discrepancy. The purpose of this paper is to describe a large body of experimental data - most of which has never been reported - that shows the effect of radial tube-related local heterogeneities

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

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.)

SPND detectors response at the control rod drop in WWER-1000. Measurement and modelling results

International Nuclear Information System (INIS)

Mitin, V.; Milto, N.; Shishkov, L.; Tsyganov, S.; Kuzmichev, M.

2006-01-01

The paper analyzes and discusses possibility of neutron flux inspection in the WWER core during fast dynamic processes applying existing in-core monitoring system. The structure and functions of the system, basic principal of detector functioning and its temporal parameters are described briefly. To assess the ability of such dynamic monitoring the event with control rod drop happened during operation of Kozloduy NPP Unit 5 is observed - at the level of power close to nominal one of the rod from control group shifted to the lowest position at-2 seconds. In-core detectors readings at the process were registered and processed with mathematical methods that allow to single out only the prompt part of the signal. Results of the processing are presented. Furthermore, the process observing have been modeled with 3D dynamic code NOSTRA. Results of modeling are presenting in a paper, and comparing with experimental ones. A good agreement achieved. The analysis of measurements and its imitation give a hope that with an aggregate signal of detectors the measurement of control rod worth could be provided, and it allows to avoid of influence of spatial effects that are significant at standard technique with ex-core ion chambers (Authors)

Simulation of leaking fuel rods

International Nuclear Information System (INIS)

Hozer, Z.

2006-01-01

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

Startup physics tests at Temelin NPP, Unit 1

International Nuclear Information System (INIS)

Sedlacek, M.; Minarcin, M.; Toth, L.; Elko, M.; Hascik, R.

2002-01-01

The objective, scope and proceedings of the physics tests of Temelin NPP, Unit 1 physical commissioning are given in this paper. Furthermore, some results of selected physics tests are presented: reactor initial criticality test, determination of reactor power range for physics testing, measurement of control rod cluster assembly group no. 10 reactivity worth in case of limitation system LS(a) actuation, control rod cluster assembly system reactivity worth measurement with single rod cluster assembly of greatest reactivity worth stuck in fully withdrawn position, measurement of differential reactivity worth of control rod cluster assembly group no. 9, boron 'endpoint' determination and measurement of power reactivity coefficient (Authors)

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.)

Neutronic Design of KALIMER-600 Core with Moderator Rods

International Nuclear Information System (INIS)

Ser Gi Hong; Sang Ji Kim; Hoon Song; Yeong Il Kim

2004-01-01

Recently, the liquid-metal reactor research team of the Korea Atomic Energy Research Institute (KAERI) designed a 600 MWe sodium-cooled, metallic fueled fast reactor meeting the goals of Generation-IV, such as economics and proliferation resistance. In this paper, the core design analysis and its performance are reported. The core is designed to have a conversion ratio slightly larger than unity with no blanket assemblies in order not to produce an excess amount of high grade plutonium and to have no need for external feeds of fissile materials. To mitigate the sodium void reactivity of the fuel-self-sufficient core with no blanket assemblies, several design changes from a reference core are tried; reduction of the active core height, annular type cores with central dummy assemblies, and the use of moderator (BeO or ZrH 2 ) rods. As a result of the analysis, it is found that of the considered designs the use of moderator rods for the softening of the core neutron spectrum is the best choice for reducing the sodium void worth with the smallest changes from the reference fuel and assembly designs. The core analysis shows that the sodium void reactivity is reduced by ∼2$ in comparison with the reference core and the core has a much more negative fuel temperature reactivity feedback in comparison with the reference core. (authors)

Hardfacing and welding rods by P/M

International Nuclear Information System (INIS)

Nayar, H.S.

1977-01-01

Certain hardfacing and welding rods are very hard and non-deformable. They are, as it is well known, generally produced by casting processes. Airco has developed a P/M process for producing these rods. The process is already practiced on a semi-production scale. In this process, the powder is poured into suitably designed and prepared molds, vibrated to pack the powder, and sintered at a temperature between the solidus and the liquidus temperatures of the alloy to produce rods with 85% or more of the theoretical density. The P/M process has some distinct advantages over the conventional casting processes. These advantages are highlighted. The process is suitable for producing Fe-, Ni-, Co-, and Cu-base hardfacing and welding rods with and without second phase hard particles such as WC. Microstructures, dimensional and density controls, weld-evaluations and hardness data are included to present evidence that the rods produced by the P/M process are suitable for various welding and hardfacing applications

Evaluation of the ability of rod drop tests to verify the stability margins in FTR

International Nuclear Information System (INIS)

Harris, R.A.; Sevenich, R.A.

1976-01-01

Predictions of the stability characteristics of FTR indicate that the reactor can be easily controlled even under the worst possible conditions. Nevertheless, experimental verification and monitoring of these characteristics will be performed during operation of the reactor. An initial evaluation of rod drop experiments which could possibly provide this verification is presented

Evaluation of the Control Rod Super Alloy Material of HTR-PM

International Nuclear Information System (INIS)

Li Pengjun; Yan He; Diao Xingzhong

2014-01-01

The control rod drive mechanism (CRDM) system is served as the first reactivity control and shutdown system for the high temperature reactor pebble-bed module (HTR-PM) in Shandong, China. And the control rod, which is pulled up and down by a chain sprocket mechanism of CRDM to realize reactivity control, compensation and shutdown, has to be durable under temperature as high as 550℃ for a long time. Thus the material persistent strength under high temperature is quite important for the reliability of the CRDM. In this paper, a review on material selection of control rod of high temperature gas cooled reactors, including AVR and THTR-300 in Germany, HTTR in Japan, PBMR in South Africa and Dragon in Britain, was summarized. The major parameters of two kinds of high temperature alloy, incoloy 800H and alloy 625, were compared and discussed. According to the ASME NH volume, a design criterion for the control rod was established and applied in the analysis of the chain by using finite element method. The numerical simulations showed that the chain made of alloy 625 could meet the condition and work for a long time under high temperature. (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.)

Polyetheretherketone (PEEK) Rods in Lumbar Spine Degenerative Disease: A Case Series.

Science.gov (United States)

Ormond, D Ryan; Albert, Ladislau; Das, Kaushik

2016-08-01

Retrospective case series. The purpose of our study was to retrospectively review the results of posterior lumbar fusion using polyetheretherketone (PEEK) rods. Pedicle screw and rod instrumentation has become the preferred technique for performing stabilization and fusion in the lumbar spine for degenerative disease. Rigid fixation with titanium rods leads to high fusion rates, but may also contribute to stress shielding and adjacent segment degeneration (ASD). Thus, some have advocated using semirigid rods made of PEEK. Although the biomechanical properties of PEEK rods have shown improved stress-shielding characteristics and anterior load-sharing properties, there are very few clinical studies evaluating their application in the lumbar spine. We evaluated a retrospective cohort of 42 patients who underwent posterior lumbar fusion from 2007 to 2009 for the treatment of lumbar spine degenerative disease using PEEK rods. Reoperation rate was the primary outcome evaluated. Fusion rate was also evaluated. Eight of the 42 patients with PEEK rods required reoperation. Reasons for reoperation mainly included ASD (5/8) and nonunion with cage migration (3/8). Radiographically, documented fusion rate was 86%. Mean follow-up was 31.4 months. No statistical differences were found in fusion rates or reoperation between age above 55 years and younger than 55 years (P=1.00), male and female (P=0.110), single or multilevel fusion (P=0.67), and fusion with and without an interbody graft (P=0.69). Smokers showed a trend towards increased risk of reoperation for ASD or instrumentation failure (P=0.056). PEEK rods demonstrate a similar fusion and reoperation rate in comparison to other instrumentation modalities in the treatment of degenerative lumbar spine disease.

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

Management of radioactive disused lightning rods

Energy Technology Data Exchange (ETDEWEB)

Santos, Paulo de Oliveira; Silva, Fabio, E-mail: pos@cdtn.br, E-mail: silvaf@cdtn.br [Centro de Desenvolvimento da Energia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2013-07-01

The manufacture of radioactive lightning rod was allowed from 1970 to 1989. This authorization was based on state-of-the art science of that time that verified that radioactive lightning rods had efficiency superior to the conventional lightning rods, denominated Franklin. However, the experience showed that their efficiency was not superior enough to justify the use of radioactive sources. Consequently, in 1989, the National Commission or Nuclear Energy - CNEN, issued the Resolution 04/89 from 04-19-1989, that forbidden the importation of {sup 241}Am tapes, assembling and commercialization of radioactive lightning-rods. The institutes of CNEN are responsible for receiving these lightning-rods and sending to the users procedures for removing and dispatch to the institutes. Therewith, these devices are kept away from the human being and environment. The Nuclear technology Development Center - CDTN and Institute for Energy and Nuclear Research - IPEN of CNEN, has built laboratories appropriate for dismantling such devices and store the {sup 241}Am tapes safely. Nowadays are being researched methodologies to evaluate the contamination levels of the frame for possible recycling and become better the management of these devices. (author)

Management of radioactive disused lightning rods

International Nuclear Information System (INIS)

Santos, Paulo de Oliveira; Silva, Fabio

2013-01-01

The manufacture of radioactive lightning rod was allowed from 1970 to 1989. This authorization was based on state-of-the art science of that time that verified that radioactive lightning rods had efficiency superior to the conventional lightning rods, denominated Franklin. However, the experience showed that their efficiency was not superior enough to justify the use of radioactive sources. Consequently, in 1989, the National Commission or Nuclear Energy - CNEN, issued the Resolution 04/89 from 04-19-1989, that forbidden the importation of 241 Am tapes, assembling and commercialization of radioactive lightning-rods. The institutes of CNEN are responsible for receiving these lightning-rods and sending to the users procedures for removing and dispatch to the institutes. Therewith, these devices are kept away from the human being and environment. The Nuclear technology Development Center - CDTN and Institute for Energy and Nuclear Research - IPEN of CNEN, has built laboratories appropriate for dismantling such devices and store the 241 Am tapes safely. Nowadays are being researched methodologies to evaluate the contamination levels of the frame for possible recycling and become better the management of these devices. (author)

Reactor core conversion studies of Ghana: Research Reactor-1 and proposal for addition of safety rod

International Nuclear Information System (INIS)

Odoi, H.C.

2014-06-01

The inclusion of an additional safety rod in conjunction with a core conversion study of Ghana Research Reactor-1 (GHARR-1) was carried out using neutronics, thermal hydraulics and burnup codes. The study is based on a recommendation by Integrated Safety Assessment for Research Reactors (INSARP) mission to incorporate a safety rod to the reactor safety system as well as the need to replace the reactor fuel with LEU. Conversion from one fuel type to another requires a complete re-evaluation of the safety analysis. Changes to the reactivity worth, shutdown margin, power density and material properties must be taken into account, and appropriate modifications made. Neutronics analysis including burnup was studied followed by thermal hydraulics analyses which comprise steady state and transients. Four computer codes were used for the analysis; MCNP, REBUS, PLTEP and PARET. The neutronics analysis revealed that the LEU core must be operated at 34 Kw in order to attain the flux of 1.0E12 n/cm 2 .s as the nominal flux of the HEU core. The auxiliary safety rod placed at a modified irradiation site gives a better worth than the cadmium capsules. For core excess reactivity of 4 mk, 348 fuel pins would be appropriate for the GHARR-1 LEU core. Results indicate that flux level of 1.0E12 n/cm 2 .s in the inner irradiation channel will not be compromised, if the power of the LEU core is increased to 34 kW. The GHARR-1 core using LEU-U0 2 -12.5% fuel can be operated for 23 shim cycles, with cycles length 2.5 years, for over 57 years at the 17 kW power level. All 23 LEU cycles meet the ∼ 4.0 mk excess reactivity required at the beginning of cycle . For comparison, the MNSR HEU reference core can also be operated for 23 shim cycles, but with a cycle length of 2.0 years for just over 46 years at 15.0kW power level. It is observed that the GHARR-1 core with LEU UO 2 fuel enriched to 12.5% and a power level of 34 kW can be operated ∼25% longer than the current HEU core operated at

Development of design technology on thermal-hydraulic performance in tight-lattice rod bundle. III - Numerical estimation on rod bowing effect based on X-ray CT data

International Nuclear Information System (INIS)

Misawa, Takeharu; Ohnuki, Akira; Katsuyama, Kozo; Nagamine, Tsuyoshi; Nakamura, Yasuo; Akimoto, Hajime; Mitsutake, Toru; Misawa, Susumu

2007-01-01

Design studies of the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) are being carried out at the Japan Atomic Energy Agency (JAEA) as one candidate for the future reactors. In actual core design, it is precondition to prevent fuel rods contact due to fuel rod bowing. However, the FLWR cores have nonconventional characteristics such as a hexagonal tight lattice arrangement and a high enrichment fuel loading. Therefore, as conservative evaluation, it is important to investigate influence of fuel rod bowing upon the boiling transition. In the JAEA, a 37-rod bundle experiments (base case test section (1.3mm gap width), gap width effect test section (1.0mm gap width), and rod bowing test section) were performed in order to investigate the thermal hydraulic characteristics in the tight lattice bundle. In this paper, the rod bowing effect test is paid attention. It is suspected that the actual fuel rod positions in the rod bowing test section may be different from the design-based positions. Even a slight displacement from the design-based position of fuel rod may occur variation of flow area, and give influence upon the thermal hydraulic characteristics in the rod bundle. Therefore, if the critical power in the rod bundle is evaluated by an analytical approach, the analysis based on more correct input can be performed by using actual fuel rod position data. In this study, the rod positions in the rod bowing test section were measured using the high energy X-ray computer tomography (Xray-CT). Based on the measured rod positions data, the subchannel analysis by the NASCA code was performed, in order to investigate applicability of the NASCA code to BT estimation of the rod bowing test section, and influence of displacement from design-based rod position upon BT estimation by the NASCA code. The predicted critical powers are agreement with those obtained by the experiment. The analysis based on the design-based rod positions is also performed, and the result is

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

Use of zero power plutonium reactor measurements as a support of criticality prediction for the SNR-300

International Nuclear Information System (INIS)

Pilate, S.; de Wouters, R.; Wehmann, U.; Helm, F.; Scholtyssek, W.

1978-01-01

Evaluations of criticality measurements performed in various SNEAK and Zero Power Plutonium Reactor (ZPPR) cores are compared. The best available methods of calculations (including transport theory) are used. The ZPPR results support well the trend indicated by the SNEAK evaluations for clean cores and for cores with followers; for cores with absorbers partially inserted, the agreement is only rough. Evaluations of control rod worth measurements are therefore also compared, using the routine method of calculation for SNR-300 (diffusion theory). The control rod worths are largely underestimated in SNEAK (C/E = 0.89), but only slightly underestimated in the ZPPR (C/E = 0.97). The difference in the nature of core fuel (uranium in SNEAK, plutonium in the ZPPR) could be at the origin of this discrepancy

Influence of implant rod curvature on sagittal correction of scoliosis deformity.

Science.gov (United States)

Salmingo, Remel Alingalan; Tadano, Shigeru; Abe, Yuichiro; Ito, Manabu

2014-08-01

Deformation of in vivo-implanted rods could alter the scoliosis sagittal correction. To our knowledge, no previous authors have investigated the influence of implanted-rod deformation on the sagittal deformity correction during scoliosis surgery. To analyze the changes of the implant rod's angle of curvature during surgery and establish its influence on sagittal correction of scoliosis deformity. A retrospective analysis of the preoperative and postoperative implant rod geometry and angle of curvature was conducted. Twenty adolescent idiopathic scoliosis patients underwent surgery. Average age at the time of operation was 14 years. The preoperative and postoperative implant rod angle of curvature expressed in degrees was obtained for each patient. Two implant rods were attached to the concave and convex side of the spinal deformity. The preoperative implant rod geometry was measured before surgical implantation. The postoperative implant rod geometry after surgery was measured by computed tomography. The implant rod angle of curvature at the sagittal plane was obtained from the implant rod geometry. The angle of curvature between the implant rod extreme ends was measured before implantation and after surgery. The sagittal curvature between the corresponding spinal levels of healthy adolescents obtained by previous studies was compared with the implant rod angle of curvature to evaluate the sagittal curve correction. The difference between the postoperative implant rod angle of curvature and normal spine sagittal curvature of the corresponding instrumented level was used to evaluate over or under correction of the sagittal deformity. The implant rods at the concave side of deformity of all patients were significantly deformed after surgery. The average degree of rod deformation Δθ at the concave and convex sides was 15.8° and 1.6°, respectively. The average preoperative and postoperative implant rod angle of curvature at the concave side was 33.6° and 17.8�

The effect of different screw-rod design on the anti-rotational torque: a biomechanical comparison of three conventional screw-rod constructs.

Science.gov (United States)

Huang, Zifang; Wang, Chongwen; Fan, Hengwei; Sui, Wenyuan; Li, Xueshi; Wang, Qifei; Yang, Junlin

2017-07-28

Screw-rod constructs have been widely used to correct spinal deformities, but the effects of different screw-rod systems on anti-rotational torque have not been determined. This study aimed to analyze the biomechanical effect of different rod-screw constructs on anti-rotational torque. Three conventional spinal screw-rod systems (Legacy, RF-F-10 and USSII) were used to test the anti-rotational torque in the material test machine. ANOVA was performed to evaluate the anti-rotational capacity of different pedicle screws-rod constructs. The anti-rotational torque of Legacy group, RF-F-10 group and USSII group were 12.3 ± 1.9 Nm, 6.8 ± 0.4 Nm, and 3.9 ± 0.8 Nm, with a P value lower than 0.05. This results indicated that the Legacy screws-rod construct could provide a highest anti-rotation capacity, which is 68% and 210% greater than RF-F-10 screw-rod construct and USSII screw-rod respectively. The anti-rotational torque may be mainly affected by screw cap and groove design. Our result showed the anti-rotational torque are: Legacy system > RF-F-10 system > USSII system, suggesting that appropriate rod-screw constructs selection in surgery may be vital for anti-rotational torque improvement and preventing derotation correction loss.

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.)

Review of control rod calibration methods for irradiated AGRs

Energy Technology Data Exchange (ETDEWEB)

Telford, A. R.R.

1975-10-15

Methods of calibrating control rods with particular reference to irradiated CAGR are surveyed. Some systematic spatial effects are found and an estimate of their magnitude made. It is concluded that control rod oscillation provides a promising method of calibrating rods at power which is as yet untried on CAGR. Also the rod drop using inverse kinetics provides a rod calibration but spatial effects may be large and these would be difficult to correct theoretically. The pulsed neutron technique provides a calibration route with small errors due to spatial effects provided a suitable K-tube can be developed. The xenon transient method is shown to have spatial effects which have not needed consideration in earlier reactors but which in CAGR would need very careful evaluation.

Fuel Rod Performance Evaluation of CE 16 x 16 LTA Operated at Steady State Using Transuranus and Pad Codes

Energy Technology Data Exchange (ETDEWEB)

Krasnorutskyy, V.; Slyeptsov, O. [Nuclear Fuel Cycle Science and Technology Establishment (NFCSTE), National Science Center, Kharkhov Institute of Physics and Technology (NSC KIPT), Kharkhov (Ukraine)

2013-03-15

The report performed under IAEA research contract No. 15370 describes the results of fuel performance evaluation of PWR fuel rods operated at steady state up to discharge burnup of {approx}60 GWD/MTU using the codes of TRANSURANUS designed by ITU and PAD designed by Westinghouse. The experimental results from US-PWR 16x16 LTA Extended Burnup Demonstration Program presented in the IFPE database of the OECD/NEA have been utilized for assessing the codes themselves during simulation of such properties as rod burnup, cladding corrosion, fuel densification and swelling, cladding irradiation growth and strain, FGR and RIP. The results obtained by PAD showed that the code properly simulates rod burnup, cladding irradiation growth and cladding oxidation with Standard Zr-4 material. The calculated burnup values along the fuel stack vary within {+-} 5% of the rod average burnup. The predicted values of the rod axial growth are (0.88-0.94) % and within the measured ones obtained in the burnup range of (50 - 60) GWD/MTU. With allowance made for probability of crud deposition and hot channel hydraulic diameter variation, the axial distribution of oxide layer is predicted well. For the nominal rod dimensions and operation conditions, the calculated peak oxide thickness is slightly overestimated based on the BE corrosion model parameters. The WEC fuel swelling and densification model together with the US NRC one, which is incorporated in the code, were used to assess the change in fuel pellet density ({Delta}{rho}) and fuel volume ({Delta}V{sub F}/V) vs. burnup as well as the rod void volume change, {Delta}V{sub V}/V, and the cladding outer diameter (OD) variation along the fuel stack. (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.)

Polyetheretherketone (PEEK) rods: short-term results in lumbar spine degenerative disease.

Science.gov (United States)

Colangeli, S; Barbanti Brodàno, G; Gasbarrini, A; Bandiera, S; Mesfin, A; Griffoni, C; Boriani, S

2015-06-01

Pedicle screw and rod instrumentation has become the preferred technique for performing stabilization and fusion in the surgical treatment of lumbar spine degenerative disease. Rigid fixation leads to high fusion rates but may also contribute to stress shielding and adjacent segment degeneration. Thus, the use of semirigid rods made of polyetheretherketone (PEEK) has been proposed. Although the PEEK rods biomechanical properties, such as anterior load sharing properties, have been shown, there are few clinical studies evaluating their application in the lumbar spine surgical treatment. This study examined a retrospective cohort of patients who underwent posterior lumbar fusion for degenerative disease using PEEK rods, in order to evaluate the clinical and radiological outcomes and the incidence of complications.

Overview of Japanese control rods development program

International Nuclear Information System (INIS)

Koyama, M.

1984-01-01

The Japanese control rods development program was established based on the fast breeder reactor program. Therefore, PNC's efforts have been made mainly for the development of analysis, design and fabrication technologies for ''JOYO'' and ''MONJU'' control rods. Laboratory studies were performed to obtain the information for absorber materials. The design and fabrication of the sealed and vented type control rod pins were completed, and water loop tests and in-sodium tests were carried out. Irradiation behavior of enriched B 4 C pellets with low and high density in DFR was examined. Japan's experimental fast reactor, JOYO, has been operated at the rated power of 50MWt and 75MWt since April 1977 when the MK-I core (breeder core) attained initial criticality. Post irradiation examinations on control rod, removed from the reactor, were carried out and their performance behavior were evaluated. In the MK-II core, a control rods monitoring program has been in investigation. Absorber Materials Irradiation Rigs (AMIR) are scheduled to be loaded and irradiated in the JOYO MK-II core from 1984. (author)

Coupled core criticality calculations with control rods located in the central reflector region

Energy Technology Data Exchange (ETDEWEB)

Sobhy, M [Reactor depatrment, nuclear research center, Inshaas (Egypt)

1995-10-01

The reactivity of a coupled core is controlled by a set of control rods distributed in the central reflector region. The reactor contains two compact cores cooled and moderated by light water. Control rods are designed to have reactivity worths sufficient to start, control and shutdown the coupled system. Each core in a coupled system is in subcritical conditions without any absorber then each core needs to the other core to fulfill nuclear chain reaction and to approach the criticality. In this case, each core is considered clean which is suitable for research reactor with low flux disturbance and better neutron economy, in addition to the advantage of disappearing the cut corner fuel baskets. This facilitate the in core fuel management with identical fuel baskets. Hot spots will disappear. This leads to a good heat transfer process. the excess reactivity and the shutdown margin are calculated for some of reflector as coupling region gives sufficient area for coupled core are calculated cost. The fluctuations of reactivity for coupled core are calculated by noise analysis technique and compared with that for rode core. The results show low reactivity perturbation associated with coupled core.

What the Common Economic Arguments against Comparable Worth Are Worth.

Science.gov (United States)

Bergmann, Barbara R.

1989-01-01

Reviews economists' views about how the economy works, from which conclusions opposing comparable worth are drawn. Discusses factors that have been omitted from economists' views--social and psychological factors that affect behavior in the workplace, permit and encourage discrimination, and have an effect on the distribution of jobs and wages.…

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.)

19 CFR 212.11 - Net worth exhibit.

Science.gov (United States)

2010-04-01

... 19 Customs Duties 3 2010-04-01 2010-04-01 false Net worth exhibit. 212.11 Section 212.11 Customs Duties UNITED STATES INTERNATIONAL TRADE COMMISSION INVESTIGATIONS OF UNFAIR PRACTICES IN IMPORT TRADE IMPLEMENTATION OF THE EQUAL ACCESS TO JUSTICE ACT Information Required From Applicants § 212.11 Net worth exhibit...

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

The turbulent flow in rod bundles

International Nuclear Information System (INIS)

Moeller, S.V.

1989-01-01

Experimental studies have shown that the axial and azimuthal turbulence intensities in the gap regions of rod bundles increase strongly with decreasing rod spacing; the fluctuating velocities in the axial and azimuthal directions have a quasi-periodic behaviour. To determine the origin of this phenomenon, an its characteristics as a function of the geometry and the Reynolds number, an experimental investigation was performed on the turbulent in several rod bundles with different aspect ratios (P/D, W/D). Hot-wires and microsphones were used for the measurements of velocity and wall pressure fluctuations. The data were evaluated to obtain spectra as well as auto and cross correlations. Based on the results, a phenomenological model is presented to explain this phenomenon. By means of the model, the mass exchange between neighbouring subchannels is explained [pt

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

A comparison of thermal algorithms of fuel rod performance code systems

International Nuclear Information System (INIS)

Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C.

2003-11-01

The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance

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

International Nuclear Information System (INIS)

Kohli, R.; Pasupathi, V.

1986-09-01

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

A comparison of thermal algorithms of fuel rod performance code systems

Energy Technology Data Exchange (ETDEWEB)

Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C

2003-11-01

The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance.

Fuel rod failure due to marked diametral expansion and fuel rod collapse occurred in the HBWR power ramp experiment

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki

1985-12-01

In the power ramp experiment with the BWR type light water loop at the HBWR, the two pre-irradiated fuel rods caused an unexpected pellet-cladding interaction (PCI). One occurred in the fuel rod with small gap of 0.10 mm, which was pre-irradiated up to the burn-up of 14 MWd/kgU. At high power, the diameter of the rod was increased markedly without accompanying significant axial elongation. The other occurred in the rod with a large gap of 0.23 mm, which was pre-irradiated up to the burn-up of 8 MWd/kgU. The diameter of the rod collapsed during a diameter measurement at the maximum power level. The causes of those were investigated in the present study by evaluating in-core data obtained from equipped instruments in the experiment. It was revealed from the investigation that these behaviours were attributed to the local reduction of the coolant flow occurred in the region of a transformer in the ramp rig. The fuel cladding material is seemed to become softened due to temperature increase caused by the local reduction of the coolant flow, and collapsed by the coolant pressure, either locally or wholly depending on the rod diametral gap existed. (author)

Measurements of negative reactivity in Masurca and Phenix control rods: Prospects for Superphenix

International Nuclear Information System (INIS)

Gauthier, J.C.; Petiot, R.; Coulon, P.; Giese, H.; West, J.P.

1986-01-01

Experimental assessment of the negative reactivity of the control rods in an industrial reactor has recently been the subject of numerous studies conducted in the light of forthcoming startup tests on the core of Superphenix. Representative tests have been carried out both on Phenix and on the Masurca critical mockup, and a test programme for Superphenix has been drawn up. Subcritical measurements (source multiplication technique) have been carried out on Phenix without absolute measurement of a standard. However, a precise relative interpretation using two counters demonstrates good agreement following the correction of spatial effects. The chief value of the rod drop measurements conducted on Masurca was that it provided a means of cross-checking the kinetic method to be validated against a standard source multiplication method. The results demonstrate complete agreement between the two methods. The acceptability of the rod drop method is therefore considered to be established. The programme foreseen for startup of Superphenix and the objectives which have been set are briefly indicated. The calculation methods to be used in respect of the startup tests have been established on the basis of experience gained through interpreting the experiments conducted in the course of the Racine (Masurca) programme. An analysis of these experiments included, among other things, a parametric study that has made it possible to devise a standard calculation method for predicting Superphenix rod worth values. The main feature is a scattering calculation with three energy groups and three dimensions. Two-dimensional scattering and transport calculations are therefore necessary in order to define the corrective factors to be applied to this initial result. The final result of this analysis is thus made equivalent to a 25-energy-group transport calculation with an extremely small spatial mesh

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.)

Evaluation of the Effect of Fixation Angle between Polyaxial Pedicle Screw Head and Rod on the Failure of Screw-Rod Connection

Directory of Open Access Journals (Sweden)

Engin Çetin

2015-01-01

Full Text Available Introduction. Polyaxial screws had been only tested according to the ASTM standards (when they were perpendicularly positioned to the rod. In this study, effects of the pedicle screws angled fixation to the rod on the mechanical properties of fixation were investigated. Materials and Method. 30 vertically fixed screws and 30 screws fixed with angle were used in the study. Screws were used in three different diameters which were 6.5 mm, 7.0 mm, and 7.5 mm, in equal numbers. Axial pull-out and flexion moment tests were performed. Test results compared with each other using appropriate statistical methods. Results. In pull-out test, vertically fixed screws, in 6.5 mm and 7.0 mm diameter, had significantly higher maximum load values than angled fixed screws with the same diameters (P<0.01. Additionally, vertically fixed screws, in all diameters, had significantly greater stiffness according to corresponding size fixed with angle (P<0.005. Conclusion. Fixing the pedicle screw to the rod with angle significantly decreased the pull-out stiffness in all diameters. Similarly, pedicle screw instrumentation fixed with angle decreased the minimum sagittal angle between the rod and the screw in all diameters for flexion moment test but the differences were not significant.

Specification of phase 3 benchmark (Hex-Z heterogeneous and burnup calculation)

International Nuclear Information System (INIS)

Kim, Y.I.

2002-01-01

During the second RCM of the IAEA Co-ordinated Research Project Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects the following items were identified as important. Heterogeneity will affect absolute core reactivity. Rod worths could be considerably reduced by heterogeneity effects depending on their detailed design. Heterogeneity effects will affect the resonance self-shielding in the treatment of fuel Doppler, steel Doppler and sodium density effects. However, it was considered more important to concentrate on the sodium density effect in order to reduce the calculational effort required. It was also recognized that burnup effects will have an influence on fuel Doppler and sodium worths. A benchmark for the assessment of heterogeneity effect for Phase 3 was defined. It is to be performed for the Hex-Z model of the reactor only. No calculations will be performed for the R-Z model. For comparison with heterogeneous evaluations, the control rod worth will be calculated at the beginning of the equilibrium cycle, based on the homogeneous model. The definitions of rod raised and rod inserted for SHR are given, using the composition numbers

Evaluation of the Behavior of Technova Corporation Rod-Stiffened Stitched Compression Specimens

Science.gov (United States)

Jegley, Dawn C.

2013-01-01

Under Space Act Agreement 1347 between NASA and Technova Corporation, Technova designed and fabricated two carbon-epoxy crippling specimens and NASA loaded them to failure in axial compression. Each specimen contained a pultruded rod stiffener which was held to the specimen skin with through-the-thickness stitches. One of these specimens was designed to be nominally the same as pultruded rod stitched specimens fabricated by Boeing under previous programs. In the other specimen, the rod was prestressed in a Technova manufacturing process to increase its ability to carrying compressive loading. Experimental results demonstrated that the specimen without prestressing carried approximately the same load as the similar Boeing specimens and that the specimen with prestressing carried significantly more load than the specimen without prestressing.

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

Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations

International Nuclear Information System (INIS)

Blomquist, R.N.; Finck, P.J.; Jammes, C.; Stenberg, C.G.

1999-01-01

Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k eff of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data

Gender Differences in Comparisons and Entitlement: Implications for Comparable Worth.

Science.gov (United States)

Major, Brenda

1989-01-01

Addresses the role of comparison processes in the persistence of the gender wage gap, its toleration by those disadvantaged by it, and resistance to comparable worth as a corrective strategy. Argues that gender segregation and undercompensation for women's jobs leads women to use different comparison standards when evaluating what they deserve.…

Characteristics of self-worth protection in achievement behaviour.

Science.gov (United States)

Thompson, T

1993-11-01

Two experiments are reported comprising an investigation of individual difference variables associated with self-worth protection. This is a phenomenon whereby students in achievement situations adopt one of a number of strategies, including withdrawing effort, in order to avoid damage to self-esteem which results from attributing failure to inability. Experiment 1 confirmed the adequacy of an operational definition which identified self-worth students on the basis of two criteria. These were deteriorated performance following failure, together with subsequent enhanced performance following a face-saving excuse allowing students to explain failure without implicating low ability. The results of Experiment 2 established that the behaviour of self-worth protective students in achievement situations may be understood in terms of their low academic self-esteem coupled with uncertainty about their level of global self-esteem. Investigation of the manner in which self-worth students explain success and failure outcomes failed to demonstrate a tendency to internalise failure but revealed a propensity on the part of these students to reject due credit for their successes. The implications of these findings in terms of the prevention and modification of self-worth protective reactions in achievement situations are discussed.

Validity of the Worth 4 Dot Test in Patients with Red-Green Color Vision Defect.

Science.gov (United States)

Bak, Eunoo; Yang, Hee Kyung; Hwang, Jeong-Min

2017-05-01

The Worth four dot test uses red and green glasses for binocular dissociation, and although it has been believed that patients with red-green color vision defects cannot accurately perform the Worth four dot test, this has not been validated. Therefore, the purpose of this study was to demonstrate the validity of the Worth four dot test in patients with congenital red-green color vision defects who have normal or abnormal binocular vision. A retrospective review of medical records was performed on 30 consecutive congenital red-green color vision defect patients who underwent the Worth four dot test. The type of color vision anomaly was determined by the Hardy Rand and Rittler (HRR) pseudoisochromatic plate test, Ishihara color test, anomaloscope, and/or the 100 hue test. All patients underwent a complete ophthalmologic examination. Binocular sensory status was evaluated with the Worth four dot test and Randot stereotest. The results were interpreted according to the presence of strabismus or amblyopia. Among the 30 patients, 24 had normal visual acuity without strabismus nor amblyopia and 6 patients had strabismus and/or amblyopia. The 24 patients without strabismus nor amblyopia all showed binocular fusional responses by seeing four dots of the Worth four dot test. Meanwhile, the six patients with strabismus or amblyopia showed various results of fusion, suppression, and diplopia. Congenital red-green color vision defect patients of different types and variable degree of binocularity could successfully perform the Worth four dot test. They showed reliable results that were in accordance with their estimated binocular sensory status.

Measurement of safety-rod effectiveness of the zero energy reactor 'RB'

Energy Technology Data Exchange (ETDEWEB)

Raisic, N; Popovic, D; Takac, S; Markovic, H; Martinc, R; Radmanovic, Lj [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

1959-03-15

The reactivity effectiveness of the two safety rods displaced eccentrically along diameter of a cylindrical D{sub 2}O moderated reactor was measured and compared with the theoretical calculations. The results show that the simplified calculations of one rod effectiveness are quite satisfactory but the theoretical evaluation of the interference effect of the two rods are not sufficiently reliable. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-07-15

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

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.)

Evaluation of the thermal-mechanical performance of fuel rods of a BWR during a power ramp using the FUELSIM code

International Nuclear Information System (INIS)

Pantoja C, R.

2010-01-01

, and the other one from 10 x 10 fuel assembly design, and a comparison of the thermal-mechanical performance between the two different rod designs is performed. Results about diverse parameters related to BWR thermal limits are presented, as maximum temperatures in the center of the fuel and results of cladding axial deformation. The performance simulations were performed by the code FUELSIM. The benefit that can be obtained from the thermal-mechanical analysis in relation to safety and economy, among others, is to design and optimize fuel rods, as well as to perform independent evaluations of the information provided by different fuel vendors. (Author)

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.)

Reactivity control system of a passively safe thorium breeder pebble bed reactor

International Nuclear Information System (INIS)

Wols, F.J.; Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der

2014-01-01

Highlights: • A worth of over 15,000 pcm ensures achieving long-term cold shutdown in thorium PBR. • Control rod worth in side reflector is insufficient due to low-power breeder zone. • 20 control rods, just outside the driver zone, can achieve long-term cold shutdown. • BF 3 gas can be inserted for reactor shutdown, but only in case of emergency. • Perturbation theory accurately predicts absorber gas worth for many concentrations. - Abstract: This work investigates the neutronic design of the reactivity control system for a 100 MW th passively safe thorium breeder pebble bed reactor (PBR), a conceptual design introduced previously by the authors. The thorium PBR consists of a central driver zone of 100 cm radius, surrounded by a breeder zone with 300 cm outer radius. The fissile content of the breeder zone is low, leading to low fluxes in the radial reflector region. Therefore, a significant decrease of the control rod worth at this position is anticipated. The reactivity worth of control rods in the side reflector and at alternative in-core positions is calculated using different techniques, being 2D neutron diffusion, perturbation theory and more accurate 3D Monte Carlo models. Sensitivity coefficients from perturbation theory provide a first indication of effective control rod positions, while the 2D diffusion models provide an upper limit on the reactivity worth achievable at a certain radial position due to the homogeneous spreading of the absorber material over the azimuthal domain. Three dimensional forward calculations, e.g. in KENO, are needed for an accurate calculation of the total control rod worth. The two dimensional homogeneous calculations indicate that the reactivity worth in the radial reflector is by far insufficient to achieve cold reactor shutdown, which requires a control rod worth of over 15 000 pcm. Three dimensional heterogeneous KENO calculations show that placing 20 control rods just outside the driver channel, between 100 cm

Reactivity control system of a passively safe thorium breeder pebble bed reactor

Energy Technology Data Exchange (ETDEWEB)

Wols, F.J., E-mail: f.j.wols@tudelft.nl; Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der

2014-12-15

Highlights: • A worth of over 15,000 pcm ensures achieving long-term cold shutdown in thorium PBR. • Control rod worth in side reflector is insufficient due to low-power breeder zone. • 20 control rods, just outside the driver zone, can achieve long-term cold shutdown. • BF{sub 3} gas can be inserted for reactor shutdown, but only in case of emergency. • Perturbation theory accurately predicts absorber gas worth for many concentrations. - Abstract: This work investigates the neutronic design of the reactivity control system for a 100 MW{sub th} passively safe thorium breeder pebble bed reactor (PBR), a conceptual design introduced previously by the authors. The thorium PBR consists of a central driver zone of 100 cm radius, surrounded by a breeder zone with 300 cm outer radius. The fissile content of the breeder zone is low, leading to low fluxes in the radial reflector region. Therefore, a significant decrease of the control rod worth at this position is anticipated. The reactivity worth of control rods in the side reflector and at alternative in-core positions is calculated using different techniques, being 2D neutron diffusion, perturbation theory and more accurate 3D Monte Carlo models. Sensitivity coefficients from perturbation theory provide a first indication of effective control rod positions, while the 2D diffusion models provide an upper limit on the reactivity worth achievable at a certain radial position due to the homogeneous spreading of the absorber material over the azimuthal domain. Three dimensional forward calculations, e.g. in KENO, are needed for an accurate calculation of the total control rod worth. The two dimensional homogeneous calculations indicate that the reactivity worth in the radial reflector is by far insufficient to achieve cold reactor shutdown, which requires a control rod worth of over 15 000 pcm. Three dimensional heterogeneous KENO calculations show that placing 20 control rods just outside the driver channel

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

Collegiate misuse of prescription stimulants: examining differences in self-worth.

Science.gov (United States)

Giordano, Amanda L; Prosek, Elizabeth A; Reader, Emily A; Bevly, Cynthia M; Turner, Kori D; LeBlanc, Yvette N; Vera, Ryan A; Molina, Citlali E; Garber, Sage Ann

2015-02-01

Prescription stimulant medication is commonly used to treat attention-deficit hyperactivity disorder (ADHD). However, stimulant medication misuse is a prevalent problem among the college population. There is limited research on psychological factors associated with collegiate nonmedical stimulant misuse. To examine the association between college students' self-worth and stimulant medication misuse. A quantitative study implemented during the 2013-2014 academic year in which we utilized a convenience sample of undergraduate students at a public university. College students (N = 3,038) completed an electronic survey packet including a stimulant use index and the Contingencies of Self-Worth Scale. We conducted descriptive discriminant analysis (DDA) to measure the associations between four groups: Nonusers, Appropriate Users, Nonprescribed Misusuers, and Prescribed Users. Significant differences in contingencies of self-worth existed between the four groups of students. Specifically, external contingencies of self-worth, such as appearance and approval, were associated with stimulant medication misuse, whereas, internal contingencies of self-worth, such as God's love and virtue, were associated with nonuse and appropriate prescribed use. Conclusions/Importance: The findings of the current study suggested contingencies of self-worth partially explain prescription stimulant misuse among the collegiate population. Addressing self-worth may be helpful in the treatment of stimulant misuse with college students.

Reliability worth assessment of radial systems with distributed generation

OpenAIRE

Bellart Llavall, Francesc Xavier

2010-01-01

With recent advances in technology, utilities generation (DG) on the distribution systems. Reliability worth is very important in power system planning and operation. Having a DG ensures reli increase the reliability worth. This research project presents the study of a radial distribution system and the impact of placing DG in order to increase the reliability worth. where a DG have to be placed. The reliability improvement is measured by different reliability indices tha...

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.)

Nuclear characteristics evaluation for Kyoto University Research Reactor with low-enriched uranium core

Energy Technology Data Exchange (ETDEWEB)

Nakajima, Ken; Unesaki, Hironobu [Kyoto University Research Reactor Institute, Kumatori-cho Sennan-gun Osaka (Japan)

2008-07-01

A project to convert the fuel of Kyoto University Research Reactor (KUR) from highly enriched uranium (HEU) to low-enriched uranium (LEU) is in progress as a part of RERTR program. Prior to the operation of LEU core, the nuclear characteristics of the core have been evaluated to confirm the safety operation. In the evaluation, nuclear parameters, such as the excess reactivity, shut down margin control rod worth, reactivity coefficients, were calculated, and they were compared with the safety limits. The results of evaluation show that the LEU core is able to satisfy the safety requirements for operation, i.e. all the parameters satisfy the safety limits. Consequently, it was confirmed that the LEU fuel core has the proper nuclear characteristics for the safety operation. (authors)

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

Managing the reactivity excess of the gas turbine-modular helium reactor by burnable poison and control rods

International Nuclear Information System (INIS)

Talamo, Alberto

2006-01-01

The gas turbine-modular helium reactor coupled to the deep burn in-core fuel management strategy offers the extraordinary capability to incinerate over 50% of the initial inventory of fissile material. This extraordinary feature, coming from an advanced and well tested fuel element design, which takes advantage of the TRISO particles technology, is maintained while the reactor is loaded with the most different types of fuels. In the present work, we assumed the reactor operating at the equilibrium of the fuel composition, obtained by a 6 years irradiation of light water reactor waste, and we investigated the effects of the introduction of the burnable poison and the control rods; we equipped the core with all the three types of control rods: operational, startup and shutdown ones. We employed as burnable poison natural erbium, due to the 167 Er increasing neutron capture microscopic cross-section in the energy range where the neutron spectrum exhibits the thermal peak; in addition, we utilized boron carbide, with 90% enrichment in 1 B, as the absorption material of the control rods. Concerning the burnable poison studies, we focused on the k eff value, the 167 Er mass during burnup, the influence of modifying the radius of the BISO particles kernel and the fuel and moderator coefficients of temperature. Concerning the control rods studies, we investigated the reactivity worth, the changes in the neutron flux profile due to a partial insertion, the influence of modifying the radius of the BISO particles kernel and the β eff , at the beginning of the operation

Managing the reactivity excess of the gas turbine-modular helium reactor by burnable poison and control rods

Energy Technology Data Exchange (ETDEWEB)

Talamo, Alberto [Department of Nuclear and Reactor Physics, Royal Institute of Technology, Roslagstullsbacken 21, S-10691, Stockholm (Sweden)]. E-mail: alby@neutron.kth.se

2006-01-15

The gas turbine-modular helium reactor coupled to the deep burn in-core fuel management strategy offers the extraordinary capability to incinerate over 50% of the initial inventory of fissile material. This extraordinary feature, coming from an advanced and well tested fuel element design, which takes advantage of the TRISO particles technology, is maintained while the reactor is loaded with the most different types of fuels. In the present work, we assumed the reactor operating at the equilibrium of the fuel composition, obtained by a 6 years irradiation of light water reactor waste, and we investigated the effects of the introduction of the burnable poison and the control rods; we equipped the core with all the three types of control rods: operational, startup and shutdown ones. We employed as burnable poison natural erbium, due to the {sup 167}Er increasing neutron capture microscopic cross-section in the energy range where the neutron spectrum exhibits the thermal peak; in addition, we utilized boron carbide, with 90% enrichment in {sup 1}B, as the absorption material of the control rods. Concerning the burnable poison studies, we focused on the k {sub eff} value, the {sup 167}Er mass during burnup, the influence of modifying the radius of the BISO particles kernel and the fuel and moderator coefficients of temperature. Concerning the control rods studies, we investigated the reactivity worth, the changes in the neutron flux profile due to a partial insertion, the influence of modifying the radius of the BISO particles kernel and the {beta} {sub eff}, at the beginning of the operation.

Expert system for control rod programming of boiling water reactors

International Nuclear Information System (INIS)

Fukuzaki, T.; Yoshida, K.; Kobayashi, Y.; Matsuura, H.; Hoshi, K.

1986-01-01

Control rod programming, one of the main tasks in reactor core management of boiling water reactors (BWRs), can be successfully accomplished by well-experienced engineers. By use of core performance evaluation codes, their knowledge plays the main role in searching through optimal control rod patterns and exposure points for adjusting notch positions and exchanging rod patterns. An expert system has been developed, based on a method of knowledge engineering, to lighten the engineer's load in control rod programming. This system utilizes an inference engine suited for planning/designing problems, and stores the knowledge of well-experienced engineers in its knowledge base. In this report, the inference engine, developed considering the characteristics of the control rod programming, is introduced. Then the constitution and function of the expert system are discussed

Pressure equalization systems in pressurized water reactor fuel rods

International Nuclear Information System (INIS)

Steven, J.; Wunderlich, F.

1979-01-01

For the development of a pressure reduction system in PWR fuel rods the capability of charcoal to adsorb Helium, Xenon and Krypton at temperatures of about 300 0 C was investigated. The influence of the adsorption on fuel rod internal pressure and in creep strain on the tube was evaluated in a design study. (orig.) [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

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.)

Formation of rod-like nanostructure by aggregation of TiO2 ...

Indian Academy of Sciences (India)

rod-like nanoparticle aggregates was evaluated by the degradation of methylene blue ... Rod-like nanostructure; aligned nanoparticle aggregates; photocatalytic activity; antibacterial ... bioactive and electrical properties by effective utilization of light. Further TiO2 ... contact with microorganism as antimicrobial nanomaterials,.

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

42 CFR 422.382 - Minimum net worth amount.

Science.gov (United States)

2010-10-01

... that CMS considers appropriate to reduce, control or eliminate start-up administrative costs. (b) After... section. (c) Calculation of the minimum net worth amount—(1) Cash requirement. (i) At the time of application, the organization must maintain at least $750,000 of the minimum net worth amount in cash or cash...

The use of routine cell codes for evaluating the In-rod effective cross sections of resonance absorber

Energy Technology Data Exchange (ETDEWEB)

Segev, M [Ben-Gurion Univ. of the Negev, Beersheba (Israel). Dept. of Nuclear Engineering

1996-12-01

The last three years have witnessed an increasing interest in the in-rod distribution of resonance absorption and of temperature. High burnup, especially beyond the `classical` limit of 30 GWd/T, is expected to generate uneven in-rod isotope distributions with consequences for fuel rod integrity and reactor Doppler feedback. There are recent indications that, even for a freshly loaded uranium-oxide rod, proper account of the U{sup 238} in-rod absorption rate distribution results in a doppler coefficient some 15% lower in magnitude than its routinely calculated value. Presently a special form of application is made of the Bogart approach. This approach is based on the fact that, as a fuel rod is filled in from the outside, its resonance capture rate increases monotonically, despite file decreasing effective capture cross section for the thickness annulus. Bogart used His observation to derive a differential equation for the in-rod absorption distribution. Presently we capitalize on the idea in a discrete form. (author).

The use of routine cell codes for evaluating the In-rod effective cross sections of resonance absorber

International Nuclear Information System (INIS)

Segev, M.

1996-01-01

The last three years have witnessed an increasing interest in the in-rod distribution of resonance absorption and of temperature. High burnup, especially beyond the 'classical' limit of 30 GWd/T, is expected to generate uneven in-rod isotope distributions with consequences for fuel rod integrity and reactor Doppler feedback. There are recent indications that, even for a freshly loaded uranium-oxide rod, proper account of the U 238 in-rod absorption rate distribution results in a doppler coefficient some 15% lower in magnitude than its routinely calculated value. Presently a special form of application is made of the Bogart approach. This approach is based on the fact that, as a fuel rod is filled in from the outside, its resonance capture rate increases monotonically, despite file decreasing effective capture cross section for the thickness annulus. Bogart used His observation to derive a differential equation for the in-rod absorption distribution. Presently we capitalize on the idea in a discrete form. (author)

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...

Analysis of the burnup of the control rods with the COREMASTER-Presto code

International Nuclear Information System (INIS)

Hernandez, J.L.; Alonso, G.; Perusquia, R.; Montes, J.L.; Hernandez, H.

2003-01-01

An evaluation of the capacity of the COREMASTER-Presto code, to evaluate generically the burnt of the control bars in the Laguna Verde reactors plant (CLV) is made. It was found that the code only reports burnt values of the control rods in MWD/TM, in spite of having with a second order polynomial model, for the conversion to remainder of the Boron-10 (B-10). It was observed that said model is adequate only for burnt smaller to 45,000 MWD/TM. To evaluate the burnt of the control rods it was reproduced the balance cycle of 18 months for the CLV, executing Cm-Presto during 13 consecutive cycles. First without rod burnt, taking this as the base case. Later on, cases with 1, 2 and up to 13 cycles with rod burnt were generated. When comparing results it was observed that the control rods pattern it loses reactivity lineally with the burnt one. By each 10 G Wd/T of burnt of the nucleus it is decreased the reactivity of the pattern rods ∼ 1 pcm in hot condition and of ∼ 20 pcm in cold condition. When burning three cycles those rods more burnt reached the 13,900 MWD/TM, equivalent to 36% of B-10 reduction, near value to 34% proposed by aging in the one lost study of B-10. It was observed that Cm-Presto it doesn't burn the superior node of the control rods when these are completely extracted. A one big lost of B-10, of the order of 50%, it represents only a decrease of 11% of the reactivity value of the rod. One can affirm that even when it is strongly decreased the content of B-10, the rod is continue considering as a black absorber, that is to say, thermal neutron that enters in the neutron rod that is absorbed. (Author)

Measurements of the Reactivity Properties of the Aagesta Nuclear Power Reactor at Zero Power

Energy Technology Data Exchange (ETDEWEB)

Bernander, G

1967-07-15

The moderator level and temperature coefficients of reactivity and control rod differential reactivity worths have been determined at zero power by means of period measurements. The moderator level coefficient and the corresponding critical level have been measured for the 32, 68 and 136 fuel assembly cores at room temperature for cores with and without control rods. From these results the worths of control rods have been derived. HETERO calculations give up to 15 % lower values than the experimental results. The cold fresh core has an excess reactivity of 9.0 {+-} 0.2 %. The temperature coefficient and differential control rod worths were measured for the fully loaded core with filled tank in the temperature range between 30 and 210 deg C. Critical positions as a function of temperature were obtained for the corresponding control rod groups. No relevant calculations of the temperature coefficient for comparison with the experimental values have yet been made, but the experimental results together with measured critical control rod positions give good opportunities to check calculational programs. HETERO has been shown in these cases to reproduce differential control rod worths and critical positions fairly well. However, a certain underestimation of the rod effectiveness is quite noticeable. The relative increase in control rod effectiveness with a temperature change from 20 to 220 deg C has been estimated to be 0.29 {+-} 0.06.

Analysis of reactivity characteristics of the MONJU initial core using JENDL-3.2

Energy Technology Data Exchange (ETDEWEB)

Sasaki, Kenji; Suzuki, Takayuki; Suzuki, Norimichi [Power Reactor and Nuclear Fuel Development Corp., Tsuruga, Fukui (Japan). Monju Construction Office; Itagaki, Yoshihiko

1998-03-01

This paper describes the evaluated results of criticality, absorber rod worth and coolant worth in the MONJU initial cores based on the JENDL-3.2 library compared with those of the JENDL-2 library. We confirm that the ratios of calculated and experimental (C/E) values using the JENDL-3.2 library are slightly better than those based on the JENDL-2 library. (author)

Spectral fine structure effects on material and doppler reactivity worth

International Nuclear Information System (INIS)

Greenspan, E.; Karni, Y.

1975-01-01

New formulations concerning the fine structure effects on the reactivity worth of resonances are developed and conclusions are derived following the extension to more general types of perturbations which include: the removal of resonance material at finite temperatures and the temperature variation of part of the resonance material. It is concluded that the flux method can overpredict the reactivity worth of resonance materials more than anticipated. Calculations on the Doppler worth were carried out; the results can be useful for asessing the contribution of the fine structure effects to the large discrepancy that exists between the calculated and measured small sample Doppler worths. (B.G.)

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.

Evaluation of oscillator strength in colloidal CdSe/CdS dots-in-rods

Energy Technology Data Exchange (ETDEWEB)

Pisanello, Ferruccio [Universite Pierre et Marie Curie, Laboratoire Kastler Brossel, CNRS UMR8552, Ecole Normale Superieure, 4 place Jussieu, 75005 Paris (France); National Nanotechnology Laboratory of CNR/INFM, Scuola superiore ISUFI, Universita del Salento, 16 Via Arnesano, 73100 Lecce (Italy); Lemenager, Godefroy; Spinicelli, Piernicola; Amo, Alberto; Giacobino, Elisabeth; Bramati, Alberto [Universite Pierre et Marie Curie, Laboratoire Kastler Brossel, CNRS UMR8552, Ecole Normale Superieure, 4 place Jussieu, 75005 Paris (France); Martiradonna, Luigi [Istituto Italiano di Tecnologia (IIT), Center for Bio-Molecular Nanotechnolgy, Via Barsanti 1, Arnesano, 73010 Lecce (Italy); Fiore, Angela [National Nanotechnology Laboratory of CNR/INFM, Scuola superiore ISUFI, Universita del Salento, 16 Via Arnesano, 73100 Lecce (Italy); Cingolani, Roberto; De Vittorio, Massimo [National Nanotechnology Laboratory of CNR/INFM, Scuola superiore ISUFI, Universita del Salento, 16 Via Arnesano, 73100 Lecce (Italy); Istituto Italiano di Tecnologia (IIT), Center for Bio-Molecular Nanotechnolgy, Via Barsanti 1, Arnesano, 73010 Lecce (Italy)

2010-11-15

The oscillator strength in CdSe/CdS colloidal dot-in-rods is evaluated and assessed to be of {proportional_to}1.5. On the basis of this finding, the possibility to reach the strong coupling regime with photonic crystals nanocavities is discussed. In spite that carefully choosing the cavity parameters the strong coupling regime could be analytically achieved at room temperature, theoretical considerations show that the typical Rabi doublet cannot be resolved. The work draws also a viable strategy toward the observation of the strong coupling at cryogenic temperatures. (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

CFD evaluation of turbulence model on heat transfer in 5 × 5 rod bundles

International Nuclear Information System (INIS)

Chao Yanmeng; Yang Lixin; Zhang Yuxiang; Pang Zhengzheng

2014-01-01

Different turbulence models may lead to different results when analyzing fuel assemblies using computational fluid dynamics (CFD) method. In this paper, a 5 × 5 rod bundle model was built to analyze the relationship between flow and heat transfer. The pressure drop and Nu were calculated using ANSYS CFX. Three factors evaluating swirling flow and cross-flow were used to analyze the inner relationship between flow field and heat transfer. The performances of various turbulence models, including eddy viscosity model and Reynold stress model, were evaluated. The comparison between numerical and similar experimental results indicates that Reynold stress model is more appropriate for modeling flow features and heat transfer in spacer grids discussed in this paper. (authors)

The biomechanical consequences of rod reduction on pedicle screws: should it be avoided?

Science.gov (United States)

Paik, Haines; Kang, Daniel G; Lehman, Ronald A; Gaume, Rachel E; Ambati, Divya V; Dmitriev, Anton E

2013-11-01

Rod contouring is frequently required to allow for appropriate alignment of pedicle screw-rod constructs. When residual mismatch is still present, a rod persuasion device is often used to achieve further rod reduction. Despite its popularity and widespread use, the biomechanical consequences of this technique have not been evaluated. To evaluate the biomechanical fixation strength of pedicle screws after attempted reduction of a rod-pedicle screw mismatch using a rod persuasion device. Fifteen 3-level, human cadaveric thoracic specimens were prepared and scanned for bone mineral density. Osteoporotic (n=6) and normal (n=9) specimens were instrumented with 5.0-mm-diameter pedicle screws; for each pair of comparison level tested, the bilateral screws were equal in length, and the screw length was determined by the thoracic level and size of the vertebra (35 to 45 mm). Titanium 5.5-mm rods were contoured and secured to the pedicle screws at the proximal and distal levels. For the middle segment, the rod on the right side was intentionally contoured to create a 5-mm residual gap between the inner bushing of the pedicle screw and the rod. A rod persuasion device was then used to engage the setscrew. The left side served as a control with perfect screw/rod alignment. After 30 minutes, constructs were disassembled and vertebrae individually potted. The implants were pulled in-line with the screw axis with peak pullout strength (POS) measured in Newton (N). For the proximal and distal segments, pedicle screws on the right side were taken out and reinserted through the same trajectory to simulate screw depth adjustment as an alternative to rod reduction. Pedicle screws reduced to the rod generated a 48% lower mean POS (495±379 N) relative to the controls (954±237 N) (p.05). In circumstances where a rod is not fully seated within the pedicle screw, the use of a rod persuasion device decreases the overall POS and work energy to failure of the screw or results in outright

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.)

Self-worth, perceived competence, and behaviour problems in children with cerebral palsy.

Science.gov (United States)

Schuengel, Carlo; Voorman, Jeanine; Stolk, Joop; Dallmeijer, Annet; Vermeer, Adri; Becher, Jules

2006-10-30

To examine the relevance of physical disabilities for self-worth and perceived competence in children with cerebral palsy (CP), and to examine associations between behaviour problems and self-worth and perceived competence. The Harter scales for self-worth and perceived competence and a new scale for perceived motor competence were used in a sample of 80 children with CP. Their motor functioning was assessed with the Gross Motor Functioning Measure (GMFM) and behaviour problems with the Child Behaviour Check List administered to parents. Self-worth and perceived competence for children with CP were comparable to the Dutch norm sample, except for perceived athletic competence. Within the CP sample, the GMFM showed a domain-specific effect on perceived motor competence. In the multivariate analysis, internalizing problems were associated negatively with all perceived competence scales and self-worth, whereas aggression was positively associated with perceived motor competence, physical appearance, and self-worth. Children with CP appear resilient against challenges posed to their self-worth caused by their disabilities. The relevance of the physical disability appears to be domain-specific. For internalizing problems and aggression, different theoretical models are needed to account for their associations with self-worth and perceived competence.

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.)

Laboratory directed research and development on disposal of plutonium recovered from weapons. FY1994 final report

International Nuclear Information System (INIS)

Pitts, J.H.; Choi, J.S.

1994-01-01

This research project was conceived as a multi-year plan to study the use of mixed plutonium oxide-uranium oxide (MOX) fuel in existing nuclear reactors. Four areas of investigation were originally proposed: (1) study reactor physics including evaluation of control rod worth and power distribution during normal operation and transients; (2) evaluate accidents focusing upon the reduced control rod worth and reduced physical properties of PuO 2 ; (3) assess the safeguards required during fabrication and use of plutonium bearing fuel assemblies; and (4) study public acceptance issues associated with using material recovered from weapons to fuel a nuclear reactor. First year accomplishments are described. Appendices contain 2 reports entitled: development and validation of advanced computational capability for MOX fueled ALWR assembly designs; and long-term criticality safety concerns associated with weapons plutonium disposition

Evolution of fuel rod support under irradiation impact on the mechanical behaviour of fuel assemblies

International Nuclear Information System (INIS)

Billerey, Antoine; Waeckel, Nicolas

2005-01-01

New fuel management targets imply to increase fuel assembly discharge burnup. Therefore, the prediction of the mechanical behaviour of the irradiated fuel assembly is essential such as excessive fuel assembly distortion induce incomplete Rod Cluster Control Assembly insertion problems (safety issue) or fuel rod vibration induced wear leading to leaking rods (plant operation problems). Within this framework, one of the most important parameter is the knowledge of the fuel rod support in the grid cell because it directly governs the mechanical behaviour of the fuel assembly and consequently allows to predict the behaviour of irradiated structures in terms of (1) axial and lateral deformation (global behaviour of the assembly) and (2) rod vibration induced wear (local behaviour of the rod). Generally, fuel rod support is provided by a spring-dimple system fixed to the grid. During irradiation, the spring force decreases and a gap between the rod and the spring may occur. This phenomenon is due to (1) stress relieving in the spring and in the dimples, (2) grid growth and (3) reduction of the rod diameter. Two models have been developed to predict the behaviour of the rod in the cell. The first model is dedicated to the evaluation of the spring force relaxation during irradiation. The second one can assess the rotation characteristic of the fuel rod in the cell, function of the spring force. The main input parameters are (1) the creep laws of the grid materials, (2) the growth law of the grid, (3) the evolution of rod diameter and (4) the design of the fuel rod support. The aim of this paper is to: (1) evaluate the consequences of grid support design modifications on the rod vibration sensitivity in terms of predicted rod to grid maximum gap during irradiation and time in operation with an open rod to grid gap, (2) evaluate, using a linear or non-linear Finite Element assembly model, the impact of the evolution of grid support under irradiation on the overall mechanical

The Consolidated Net Worth of Private Colleges. Recommendation of a Model.

Science.gov (United States)

Jenny, Hans H.

One of several essential tools for assessing how the financial health of educational institutions is evolving is the Consolidated Net Worth Statement. This essay explores various aspects of conventional "funds" balance sheets and compares them with the Consolidated Net Worth. Emphasis is placed on how the Consolidated Net Worth Statement…

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

Evolution of fuel rod support under irradiation consequences on the mechanical behavior of fuel assembly

International Nuclear Information System (INIS)

Billerey, A.; Bouffioux, P.

2002-01-01

The complete paper follows. According to the fuel management policy in French PWR with respect to high burn-up, the prediction of the mechanical behavior of the irradiated fuel assembly is required as far as excessive deformations of fuel assembly might lead to incomplete Rod Cluster Control Assembly insertion (safety problems) and fretting wear lead to leaking rods (plant operation problems). One of the most important parameter is the evolution of the fuel rod support in the grid cell as it directly governs the mechanical behavior of the fuel assembly and consequently allows to predict the behavior of irradiated structure in terms of (i) axial and lateral deformation (global behavior of the assembly) and (ii) fretting wear (local behavior of the rod). Fuel rod support is provided by a spring-dimple system fixed on the grid. During irradiation, the spring force decreases and a gap between the rod and the spring might open. This phenomenon is due to (i) irradiation-induced stress relaxation for the spring and for the dimples, (ii) grid growth and (iii) reduction of rod diameter. Two models have been developed to predict the behavior of the rod in the grid cell. The first model is able to evaluate the spring force relaxation during irradiation. The second one is able to evaluate the rotation characteristic of the fuel rod in the cell, function of the spring force. The main input parameters are (i) the creep laws of the grid materials, (ii) the growth law of the grid, (iii) the evolution of rod diameter and (iv) the design of the fuel rod support. The objectives of this paper are to: (i) evaluate the consequences of grid support design modifications on the fretting sensitivity in terms of predicted maximum gap during irradiation and operational time to gap appearance; (ii) evaluate, using a non-linear Finite Element assembly model, the impact of the evolution of grid support under irradiation on the mechanical behavior of the full assembly in terms of axial and

Evaluation and planning for lightning rod grounding of PSTA cyclotron building

International Nuclear Information System (INIS)

Suyamto; Taufik; Idrus Abdul Kudus

2015-01-01

Lightning rod connected with the ground resistance is an equipment protection against hazards of lightning strikes building. Lightning strike to the building may result in damage to the building and destroy all the equipment inside it. The need for a lightning rod of a building is regulated in PUIPP expressed with risk factors (FR). The amount of FR is the sum of the value of the index of five (5 ) components of the building i.e building functions, construction, the height and the situation of the building and and the number of yearly lightning days in that places. At this time 05 PSTA building has undergone changes in the function of the building's mechanical workshop into a cyclotron building so that safety criteria also change into vital building with lightning rods resistance have to < 1 Ω. From measurements of grounding resistant which exist at present known that average Rp is 1.26 Ω so it is necessary to install new additional grounding resistance to reduce being less than 1 Ω. To fulfil this and taking into consideration the cost and ease of installation, planned addition of a grounding using electrodes solid rods of copper, a diameter of 16 mm and a length of 4 m , planted the soil water depth of 12 m, as well as clay covering, with a water content of about 30 %. Under these conditions and taking into the cost and ease of installation are expected to obtain optimal results i.e. soil resistivity 18.35 Ω-m and its resistance of Rx 4.82 Ω. When coupled with existing grounding final resistant Rp 0.99 Ω obtained is thus fulfilling the requirements of PUIPP that is less than 1 Ω. (author)

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.)

Calculation code of heterogeneity effects for analysis of small sample reactivity worth

International Nuclear Information System (INIS)

Okajima, Shigeaki; Mukaiyama, Takehiko; Maeda, Akio.

1988-03-01

The discrepancy between experimental and calculated central reactivity worths has been one of the most significant interests for the analysis of fast reactor critical experiment. Two effects have been pointed out so as to be taken into account in the calculation as the possible cause of the discrepancy; one is the local heterogeneity effect which is associated with the measurement geometry, the other is the heterogeneity effect on the distribution of the intracell adjoint flux. In order to evaluate these effects in the analysis of FCA actinide sample reactivity worth the calculation code based on the collision probability method was developed. The code can handle the sample size effect which is one of the local heterogeneity effects and also the intracell adjoint heterogeneity effect. (author)

Safety analysis of control rod drive computers

International Nuclear Information System (INIS)

Ehrenberger, W.; Rauch, G.; Schmeil, U.; Maertz, J.; Mainka, E.U.; Nordland, O.; Gloee, G.

1985-01-01

The analysis of the most significant user programmes revealed no errors in these programmes. The evaluation of approximately 82 cumulated years of operation demonstrated that the operating system of the control rod positioning processor has a reliability that is sufficiently good for the tasks this computer has to fulfil. Computers can be used for safety relevant tasks. The experience gained with the control rod positioning processor confirms that computers are not less reliable than conventional instrumentation and control system for comparable tasks. The examination and evaluation of computers for safety relevant tasks can be done with programme analysis or statistical evaluation of the operating experience. Programme analysis is recommended for seldom used and well structured programmes. For programmes with a long, cumulated operating time a statistical evaluation is more advisable. The effort for examination and evaluation is not greater than the corresponding effort for conventional instrumentation and control systems. This project has also revealed that, where it is technologically sensible, process controlling computers or microprocessors can be qualified for safety relevant tasks without undue effort. (orig./HP) [de

29 CFR 4062.4 - Determinations of net worth and collective net worth.

Science.gov (United States)

2010-07-01

... financial condition, and business history. (6) The economic outlook for the person's industry and the market... do not produce income for the business being valued or are not used in the business. (c) Factors for... to sell, or offer to purchase or sell the business of the person made on or about the net worth...

Preliminary physics calculations for the Clinch River Breeder Reactor

International Nuclear Information System (INIS)

Kalimullah.

1975-01-01

Calculations of sodium void, fuel, and clad worths, power distribution, and control rod worths have been carried out for an R-Z model of the CRBR, using diffusion theory and first-order perturbation theory for material worths. The power distribution and control rod worths have also been calculated in two-dimensional triangular mesh geometry. The present results are preliminary because of inaccuracy of the reactor model and the cross sections used, but the final results are not expected to be greatly different. (U.S.)

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.

MCNP evaluation of top node control rod depletion below the core in KKL

International Nuclear Information System (INIS)

Beran, Tâm; Seltborg, Per; Lindahl, Sten-Örjan; Bieli, Roger; Ledergerber, Guido

2014-01-01

In previous studies, there has been identified a significant discrepancy in the BWR control rod top node depletion between the two core simulator nodal codes POLCA7 and PRESTO-2, which indicates that there is a large general uncertainty in nodal codes in calculating the top node depletion of fully withdrawn control rods. In this study, the stochastic Monte Carlo code MCNP has been used to calculate the top node control rod depletion for benchmarking the nodal codes. By using the TIP signal obtained from an extended TIP campaign below the core performed in the KKL reactor, the MCNP model has been verified by comparing the axial profile between the TIP data and the gamma flux calculated by MCNP. The MCNP results have also been compared with calculations from POLCA7, which was found to yield slightly higher depletion rates than MCNP. It was also found that the 10 B depletion in the top node is very sensitive to the exact axial location of the control rod top when it is fully withdrawn. By using the MCNP results, the neutron flux model below the core in the nodal codes can be improved by implementing an exponential function for the neutron flux. (author)

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.)

Evaluation of the Westinghouse 10B depletion for BWR control rods

International Nuclear Information System (INIS)

Vallgren, Christina

2008-03-01

The aim of this work was to establish the 10 B depletion model for CR 99 control rods by using the latest version of POLCA7. In order to obtain an understanding of the differences between the currently used 10 B depletion models implemented in POLCA4 at O3 and in SIMULATE-3 at OL1, and the latest improved model implemented in the latest POLCA7, this work has been performed in three different parts. The first part of the work was to find out how large differences there exist in 10 B depletion between the calculated data by using the latest core monitoring system (POLCA7 version 4.10.0) and the measured data obtained in the hot-cell laboratory in Studsvik. It was found that the 10 B depletion computed by the latest POLCA7 version is in good agreement with the measured data from Studsvik. A poor agreement with a conservative overestimation in 10 B depletion was also found between the old model and the measured data. The aim of the second part of the work was to compare the calculated 10 B depletion values for two CR 99 rods from Olkiluoto 1 with the calculated 10 B depletion value for a CR 99 rod from Oskarshamn 3, by using the new 10 B depletion model implemented in the latest POLCA7 version. Swelling measurements of the boron carbide pins, used as absorber material, have indicated that the 10 B depletion should be of similar magnitude for the rods in Olkiluoto 1 and the rod in Oskarshamn 3, whereas the calculated values by using the earlier 10 B depletion models on the process computers showed a difference of about 20 %. By using the new 10 B depletion model m POLCA7, it was found that the 10 B depletion in the two studied cases was similar to each other and, thus, the hypothesis of a linear relationship between B 4 C swelling and thermal neutron fluence was supported. This third part of the work was carried out at KKL, Switzerland, and focused on comparing the B depletion models used in Westinghouse/POLCA7 and KKL/PRESTO-2. It was found that there is a slight

Evaluation of the Westinghouse 10B depletion for BWR control rods

Energy Technology Data Exchange (ETDEWEB)

Vallgren, Christina

2008-03-15

The aim of this work was to establish the 10B depletion model for CR 99 control rods by using the latest version of POLCA7. In order to obtain an understanding of the differences between the currently used 10B depletion models implemented in POLCA4 at O3 and in SIMULATE-3 at OL1, and the latest improved model implemented in the latest POLCA7, this work has been performed in three different parts. The first part of the work was to find out how large differences there exist in 10B depletion between the calculated data by using the latest core monitoring system (POLCA7 version 4.10.0) and the measured data obtained in the hot-cell laboratory in Studsvik. It was found that the 10B depletion computed by the latest POLCA7 version is in good agreement with the measured data from Studsvik. A poor agreement with a conservative overestimation in 10B depletion was also found between the old model and the measured data. The aim of the second part of the work was to compare the calculated 10B depletion values for two CR 99 rods from Olkiluoto 1 with the calculated 10B depletion value for a CR 99 rod from Oskarshamn 3, by using the new 10B depletion model implemented in the latest POLCA7 version. Swelling measurements of the boron carbide pins, used as absorber material, have indicated that the 10B depletion should be of similar magnitude for the rods in Olkiluoto 1 and the rod in Oskarshamn 3, whereas the calculated values by using the earlier 10B depletion models on the process computers showed a difference of about 20 %. By using the new 10B depletion model m POLCA7, it was found that the 10B depletion in the two studied cases was similar to each other and, thus, the hypothesis of a linear relationship between B{sub 4}C swelling and thermal neutron fluence was supported. This third part of the work was carried out at KKL, Switzerland, and focused on comparing the B depletion models used in Westinghouse/POLCA7 and KKL/PRESTO-2. It was found that there is a slight difference in

Comparison of diffusion and transport theory analysis with experimental results in fast breeder test reactor

International Nuclear Information System (INIS)

Sathyabama, N.; Mohanakrishnan, P.; Lee, S.M.

1994-01-01

A systematic analysis has been performed by 3 dimensional diffusion and transport methods to calculate the measured control rod worths and subassembly wise power distribution in fast breeder test reactor. Geometry corrections (rectangular to hexagonal and diffusion to transport corrections are estimated for multiplication factors and control rod worths. Calculated control rod worths by diffusion and transport theory are nearly the same and 10% above measured values. Power distribution in the core periphery is over predicted (15%) by diffusion theory. But, this over prediction reduces to 8% by use of the S N method. (authors). 9 refs., 4 tabs., 3 fig

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)

Associations of adolescent hopelessness and self-worth with pregnancy attempts and pregnancy desire.

Science.gov (United States)

Fedorowicz, Anna R; Hellerstedt, Wendy L; Schreiner, Pamela J; Bolland, John M

2014-08-01

We examined the associations of pregnancy desire (ambivalence or happiness about a pregnancy in the next year) and recent pregnancy attempts with hopelessness and self-worth among low-income adolescents. To evaluate independent associations among the study variables, we conducted gender-stratified multivariable logistic regression analyses with data derived from 2285 sexually experienced 9- to 18-year-old participants in the Mobile Youth Survey between 2006 and 2009. Fifty-seven percent of youths reported a desire for pregnancy and 9% reported pregnancy attempts. In multivariable analyses, hopelessness was positively associated and self-worth was negatively associated with pregnancy attempts among both female and male youths. Hopelessness was weakly associated (P = .05) with pregnancy desire among female youths. The negative association of self-worth and the positive association of hopelessness with pregnancy attempts among young men as well as young women and the association of hopelessness with pregnancy desire among young women raise questions about why pregnancy is apparently valued by youths who rate their social and cognitive competence as low and who live in an environment with few options for material success.

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.)

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

Investigation of axial power gradients near a control rod tip

Energy Technology Data Exchange (ETDEWEB)

Loberg, John, E-mail: John.Loberg@fysast.uu.se [Uppsala University, Department of Physics and Astronomy, Division of Applied Nuclear Physics, Box 525, SE-75120 Uppsala (Sweden); Osterlund, Michael, E-mail: Michael.Osterlund@fysast.uu.se [Uppsala University, Department of Physics and Astronomy, Division of Applied Nuclear Physics, Box 525, SE-75120 Uppsala (Sweden); Bejmer, Klaes-Hakan, E-mail: Klaes-Hakan.Bejmer@vattenfall.com [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, 162 60 Vaellingby, Stockholm (Sweden); Blomgren, Jan, E-mail: Jan.Blomgren@vattenfall.com [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, 162 60 Vaellingby, Stockholm (Sweden); Kierkegaard, Jesper, E-mail: Jesper.Kierkegaar@vattenfall.com [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, 162 60 Vaellingby, Stockholm (Sweden)

2011-07-15

Highlights: > Pin power gradients near BWR control rod tips have been investigated. > A control rod tip is modeled in MCNP and compared to simplified 2D/3D geometry. > Small nodes increases pin power gradients; standard nodes underestimates gradients. > The MCNP results are validated against axial gamma scan of a controlled fuel pin. - Abstract: Control rod withdrawal in BWRs induces large power steps in the adjacent fuel assemblies. This paper investigates how well a 2D/3D method, e.g., CASMO5/SIMULATE5 computes axial pin power gradients adjacent to an asymmetrical control-rod tip in a BWR. The ability to predict pin power gradients accurately is important for safety considerations whereas large powers steps induced by control rod withdrawal can cause Pellet Cladding Interaction. The computation of axial pin power gradients axially around a control rod tip is a challenging task for any nodal code. On top of that, asymmetrical control rod handles are present in some BWR designs. The lattice code CASMO requires diagonal symmetry of all control rod parts. This introduces an error in computed pin power gradients that has been evaluated by Monte Carlo calculations. The results show that CASMO5/SIMULATE5, despite the asymmetrical control rod handle, is able to predict the axial pin power gradient within 1%/cm for axial nodal sizes of 15-3.68 cm. However, a nodal size of 3.68 cm still causes underestimations of pin power gradients compared with 1 cm nodes. Furthermore, if conventional node sizes are used, {approx}15 cm, pin power gradients can be underestimated by over 50% compared with 1 cm nodes. The detailed axial pin power profiles from MCNP are corroborated by measured gamma scan data on fuel rods irradiated adjacent to control rods.

Investigation of axial power gradients near a control rod tip

International Nuclear Information System (INIS)

Loberg, John; Osterlund, Michael; Bejmer, Klaes-Hakan; Blomgren, Jan; Kierkegaard, Jesper

2011-01-01

Highlights: → Pin power gradients near BWR control rod tips have been investigated. → A control rod tip is modeled in MCNP and compared to simplified 2D/3D geometry. → Small nodes increases pin power gradients; standard nodes underestimates gradients. → The MCNP results are validated against axial gamma scan of a controlled fuel pin. - Abstract: Control rod withdrawal in BWRs induces large power steps in the adjacent fuel assemblies. This paper investigates how well a 2D/3D method, e.g., CASMO5/SIMULATE5 computes axial pin power gradients adjacent to an asymmetrical control-rod tip in a BWR. The ability to predict pin power gradients accurately is important for safety considerations whereas large powers steps induced by control rod withdrawal can cause Pellet Cladding Interaction. The computation of axial pin power gradients axially around a control rod tip is a challenging task for any nodal code. On top of that, asymmetrical control rod handles are present in some BWR designs. The lattice code CASMO requires diagonal symmetry of all control rod parts. This introduces an error in computed pin power gradients that has been evaluated by Monte Carlo calculations. The results show that CASMO5/SIMULATE5, despite the asymmetrical control rod handle, is able to predict the axial pin power gradient within 1%/cm for axial nodal sizes of 15-3.68 cm. However, a nodal size of 3.68 cm still causes underestimations of pin power gradients compared with 1 cm nodes. Furthermore, if conventional node sizes are used, ∼15 cm, pin power gradients can be underestimated by over 50% compared with 1 cm nodes. The detailed axial pin power profiles from MCNP are corroborated by measured gamma scan data on fuel rods irradiated adjacent to control rods.

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

Proceedings of the first analysis meeting on JUPITER Program. Report SN-241-80-14

Energy Technology Data Exchange (ETDEWEB)

NONE

1980-09-01

These proceedings provide an analysis of the JUPITER Experiment in ZPPR-9. A general description of data and methods is provided along with discussions of criticality, reaction rate ratio and reaction rate, Doppler and sample reactivity worth, sodium void worth, and control rod worth. Topics provided are: Considerations on a Critical Experiment Program for a Large Fast Breeder Reactor; Analysis of Sodium Void Worths in ZPPR-3 Modified Phase 3 Core; Fuel Pin and Subassembly Heterogeneity Effect on Neutronics Properties of a Fast Power Reactor; and An Evaluation Method for Mesh Size Effects in Neutron Diffusion Calculations. The individual topics have been cataloged separately.

Self-worth, perceived competence, and behaviour problems in children with cerebral palsy

NARCIS (Netherlands)

Schuengel, C.; Voorman, J.; Stolk, J.; Dallmeijer, A.J.; Vermeer, A; Becher, J.

2006-01-01

Purpose. To examine the relevance of physical disabilities for self-worth and perceived competence in children with cerebral palsy (CP), and to examine associations between behaviour problems and self-worth and perceived competence. Methods. The Harter scales for self-worth and perceived competence

The nuclear fuel rod character recognition system based on neural network technique

International Nuclear Information System (INIS)

Kim, Woong-Ki; Park, Soon-Yong; Lee, Yong-Bum; Kim, Seung-Ho; Lee, Jong-Min; Chien, Sung-Il.

1994-01-01

The nuclear fuel rods should be discriminated and managed systematically by numeric characters which are printed at the end part of each rod in the process of producing fuel assembly. The characters are used to examine manufacturing process of the fuel rods in the inspection process of irradiated fuel rod. Therefore automatic character recognition is one of the most important technologies to establish automatic manufacturing process of fuel assembly. In the developed character recognition system, mesh feature set extracted from each character written in the fuel rod is employed to train a neural network based on back-propagation algorithm as a classifier for character recognition system. Performance evaluation has been achieved on a test set which is not included in a training character set. (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.)

Exploring the relationship between appearance-contingent self-worth and self-esteem: The roles of self-objectification and appearance anxiety.

Science.gov (United States)

Adams, Katherine E; Tyler, James M; Calogero, Rachel; Lee, Jenifer

2017-12-01

Previous work has shown that both an appearance-contingent self-worth (i.e., staking one's overall self-evaluation on one's physical appearance) and self-objectification are associated with higher appearance anxiety and lower self-esteem among women. Although prior evidence separately links both appearance-contingent self-worth and self-objectification to these negative outcomes, no work has examined the mediating processes that may underlie this relationship. With the current project, we examined the relationship between appearance-contingent self-worth and self-objectification, and the degree to which this relationship is associated with higher appearance anxiety and lower overall self-esteem. We hypothesized that appearance-contingent self-worth would be positively associated with self-objectification; in turn, we expected self-objectification to be related to higher appearance anxiety, and ultimately, lower self-esteem. Across two studies, one cross-sectional (N=208) and one short-term longitudinal (N=191), we found compelling support for this hypothesis. These findings have practical and theoretical significance for both the self-objectification and contingent self-worth literatures. Published by Elsevier Ltd.

Reducing contingent self-worth: a defensive response to self-threats.

Science.gov (United States)

Buckingham, Justin; Lam, Tiffany A; Andrade, Fernanda C; Boring, Brandon L; Emery, Danielle

2018-04-10

Previous research shows that people with high self-esteem cope with threats to the self by reducing the extent to which their self-worth is contingent on the threatened domain (Buckingham, Weber, & Sypher, 2012). The present studies tested the hypothesis that this is a defensive process. In support of this hypothesis, Study 1 (N = 160), showed that self-affirmation attenuates the tendency for people with high self-esteem to reduce their contingencies of self-worth following self-threat. Furthermore, Study 2 (N = 286), showed that this tendency was more prevalent among people with defensive self-esteem than among those with secure self-esteem. The present studies imply that reducing contingent self-worth after self-threat is a defensive process. We discuss implications for theories of contingent self-worth.

Modelling of pellet cladding interaction during power ramps in PWR rods by means of Transuranus fuel rod analysis code

International Nuclear Information System (INIS)

Di Marcello, V.; Luzzi, L.

2008-01-01

Pellet-cladding interaction (PCI) in PWR type rods subjected to power ramps was analysed by means of TRANSURANUS (TU) fuel rod performance code. PCI phenomena depend on the fuel power history - i.e. by several irradiation and thermal induced phenomena occurring in the fuel rod and mutually interacting during its life in reactor - and may become critical for cladding integrity under accidental conditions. Ten test fuel rods, whose power histories and post irradiation experiment (PIE) data were available from the OECD/NEA-IAEA International Fuel Performance Experiment (UTE) database through the Studsvik SUPER-RAMP Project, were simulated by TRANSURANUS. During a power ramp pellet gaseous swelling can be inhibited by cladding pressure and can be over-predicted by a normal operation swelling model. This phenomenon was simulated by a new formulation of a fuel swelling model already available in the code, in order to consider hot pressing of inter-granular -fuel porosity due to the high hydrostatic stress resulting from PCI: it was found that TRANSURANUS, as a result of the proposed swelling formulation as well as of the accurate modelling of the other phenomena occurring during irradiation, gives correct predictions on PCI induced fuel rod failures. In addition, PCI failure threshold identified by TRANSURANUS was compared with the technological limits known in literature: the possibility of relaxing these limits for low burn-up values and the preponderance of the European fuel rod design in front of PCI emerged from TU analyses. Finally, a good agreement was found between TU evaluations and PIE data, with regard to fission gas release, fuel grain growth, and creep, corrosion and elongation of the cladding. (authors)

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

Mechanical stress analysis for a fuel rod under normal operating conditions

International Nuclear Information System (INIS)

Pino, Eddy S.; Giovedi, Claudia; Serra, Andre da Silva; Abe, Alfredo Y.

2013-01-01

Nuclear reactor fuel elements consist mainly in a system of a nuclear fuel encapsulated by a cladding material subject to high fluxes of energetic neutrons, high operating temperatures, pressure systems, thermal gradients, heat fluxes and with chemical compatibility with the reactor coolant. The design of a nuclear reactor requires, among a set of activities, the evaluation of the structural integrity of the fuel rod submitted to different loads acting on the fuel rod and the specific properties (dimensions and mechanical and thermal properties) of the cladding material and coolant, including thermal and pressure gradients produced inside the rod due to the fuel burnup. In this work were evaluated the structural mechanical stresses of a fuel rod using stainless steel as cladding material and UO 2 with a low degree of enrichment as fuel pellet on a PWR (pressurized water reactor) under normal operating conditions. In this sense, tangential, radial and axial stress on internal and external cladding surfaces considering the orientations of 0 deg, 90 deg and 180 deg were considered. The obtained values were compared with the limit values for stress to the studied material. From the obtained results, it was possible to conclude that, under the expected normal reactor operation conditions, the integrity of the fuel rod can be maintained. (author)

Quantifying data worth toward reducing predictive uncertainty

Science.gov (United States)

Dausman, A.M.; Doherty, J.; Langevin, C.D.; Sukop, M.C.

2010-01-01

The present study demonstrates a methodology for optimization of environmental data acquisition. Based on the premise that the worth of data increases in proportion to its ability to reduce the uncertainty of key model predictions, the methodology can be used to compare the worth of different data types, gathered at different locations within study areas of arbitrary complexity. The method is applied to a hypothetical nonlinear, variable density numerical model of salt and heat transport. The relative utilities of temperature and concentration measurements at different locations within the model domain are assessed in terms of their ability to reduce the uncertainty associated with predictions of movement of the salt water interface in response to a decrease in fresh water recharge. In order to test the sensitivity of the method to nonlinear model behavior, analyses were repeated for multiple realizations of system properties. Rankings of observation worth were similar for all realizations, indicating robust performance of the methodology when employed in conjunction with a highly nonlinear model. The analysis showed that while concentration and temperature measurements can both aid in the prediction of interface movement, concentration measurements, especially when taken in proximity to the interface at locations where the interface is expected to move, are of greater worth than temperature measurements. Nevertheless, it was also demonstrated that pairs of temperature measurements, taken in strategic locations with respect to the interface, can also lead to more precise predictions of interface movement. Journal compilation ?? 2010 National Ground Water Association.

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

Calculation of isotope burn-up and change in efficiency of absorbing elements of WWER-1000 control and protection system during burn-up

International Nuclear Information System (INIS)

Timofeeva, O.A.; Kurakin, K.U.

2006-01-01

The report deals with fast and thermal neutron flows distribution in structural elements of WWER-1000 fuel assembly and absorbing rods, determination of absorbing isotope burn-up and worth variation in WWER reactor control and protection system rods. Simulation of absorber rod burn-up is provided using code package SAPPHIRE 9 5 end RC W WER allowing detailed description of the core segment spatial model. Maximum burn-up of absorbing rods and respective worth variation of control and protection system rods is determined on the basis of a number of calculations considering known characteristics of fuel cycles (Authors)

Knowledge based system for control rod programming of BWRs

International Nuclear Information System (INIS)

Fukuzaki, Takaharu; Yoshida, Ken-ichi; Kobayashi, Yasuhiro

1988-01-01

A knowledge based system has been developed to support designers in control rod programming of BWRs. The programming searches through optimal control rod patterns to realize safe and effective burning of nuclear fuel. Knowledge of experienced designers plays the main role in minimizing the number of calculations by the core performance evaluation code. This code predicts power distibution and thermal margins of the nuclear fuel. This knowledge is transformed into 'if-then' type rules and subroutines, and is stored in a knowledge base of the knowledge based system. The system consists of working area, an inference engine and the knowledge base. The inference engine can detect those data which have to be regenerated, call those subroutine which control the user's interface and numerical computations, and store competitive sets of data in different parts of the working area. Using this system, control rod programming of a BWR plant was traced with about 500 rules and 150 subroutines. Both the generation of control rod patterns for the first calculation of the code and the modification of a control rod pattern to reflect the calculation were completed more effectively than in a conventional method. (author)

Fuel rod design by statistical methods for MOX fuel

International Nuclear Information System (INIS)

Heins, L.; Landskron, H.

2000-01-01

Statistical methods in fuel rod design have received more and more attention during the last years. One of different possible ways to use statistical methods in fuel rod design can be described as follows: Monte Carlo calculations are performed using the fuel rod code CARO. For each run with CARO, the set of input data is modified: parameters describing the design of the fuel rod (geometrical data, density etc.) and modeling parameters are randomly selected according to their individual distributions. Power histories are varied systematically in a way that each power history of the relevant core management calculation is represented in the Monte Carlo calculations with equal frequency. The frequency distributions of the results as rod internal pressure and cladding strain which are generated by the Monte Carlo calculation are evaluated and compared with the design criteria. Up to now, this methodology has been applied to licensing calculations for PWRs and BWRs, UO 2 and MOX fuel, in 3 countries. Especially for the insertion of MOX fuel resulting in power histories with relatively high linear heat generation rates at higher burnup, the statistical methodology is an appropriate approach to demonstrate the compliance of licensing requirements. (author)

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.)

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)

INDIAN POINT REACTOR STARTUP AND PERFORMANCE

Energy Technology Data Exchange (ETDEWEB)

Deddens, J. C.; Batch, M. L.

1963-09-15

The testing program for the Indian Point Reactor is discussed. The thermal and hydraulic evaluation of the primary coolant system is discussed. Analyses of fuel loading and initial criticality, measurement of operating coefficients of reactivity, control rod group reactivity worths, and xenon evaluation are presented. (R.E.U.)

Assessment of the MDNBR enhancement methodologies for the SMART control rods banks withdrawal event

International Nuclear Information System (INIS)

Yang, Soo Hyung; Chung, Young-Jong; Kim, Hee Cheol

2005-01-01

For an electricity generation and seawater desalination, a 330 MW System-integrated Modular Advanced ReacTor (SMART) was developed by KAERI. The safety level of the SMART is enhanced when compared to that of the typical commercial reactors, with the aid of an elimination of a large break loss of coolant accident by placing the major components of the primary system in a reactor vessel and the adoption of a new technology and a passive design concept into the safety system. However, the events related to reactivity and power distribution anomalies have been evaluated as vulnerable points when compared to the other initiating events in the SMART, since the reactivity worth of the control rods (CR) banks is quite large due to the boron free core concept. Especially, safety margins, i.e., minimum departure from nucleate boiling ratio (MDNBR), are significantly threatened during the CR banks withdrawal event. Therefore, MDNBR enhancement methodology for the CR banks withdrawal event should be considered to further enhance the safety level of the SMART design. Two methodologies have been suggested to enhance the MDNBR during the CR banks withdrawal event: the application of a DNBR trip function into a core protection system and a turbine trip delay methodology. Sensitivity studies are performed to evaluate the two MDNBR enhancement methodologies and show that the suggested methodologies could enhance the MDNBR during the CR banks withdrawal event of the SMART

Achieving equal pay for comparable worth through arbitration.

Science.gov (United States)

Wisniewski, S C

1982-01-01

Traditional "women's jobs" often pay relatively low wages because of the effects of institutionalized stereotypes concerning women and their role in the work place. One way of dealing with sex discrimination that results in job segregation is to narrow the existing wage differential between "men's jobs" and "women's jobs." Where the jobs are dissimilar on their face, this narrowing of pay differences involves implementing the concept of "equal pay for jobs of comparable worth." Some time in the future, far-reaching, perhaps even industrywide, reductions in male-female pay differentials may be achieved by pursuing legal remedies based on equal pay for comparable worth. However, as the author demonstrates, immediate, albeit more limited, relief for sex-based pay inequities found in specific work places can be obtained by implementing equal pay for jobs of comparable worth through the collective bargaining and arbitration processes.

CHF prediction in rod bundles using round tube data

Energy Technology Data Exchange (ETDEWEB)

Souza, Wallen F.; Veloso, Maria A.F.; Pereira, Cláubia; Costa, Antonella L., E-mail: wallenfds@yahoo.com.br, E-mail: mdora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-07-01

The present work concerns the use of 1995 CHF table for uniformly heated round tubes, developed jointly by Canadian and Russian researchers, for the prediction of critical heat fluxes in rod bundles geometries. Comparisons between measured and calculated critical heat fluxes indicate that this table could be applied to rod bundles provided that a suitable correction factor is employed. The tolerance limits associated with the departure from nucleate boiling ratio (DNBR) are evaluated by using statistical analysis. (author)

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.)

Criticality experiment for No.2 core of DF-VI fast neutron criticality facility

International Nuclear Information System (INIS)

Yang Lijun; Liu Zhenhua; Yan Fengwen; Luo Zhiwen; Chu Chun; Liang Shuhong

2007-01-01

At the completion of the DF-VI fast neutron criticality facility, its core changed, and it was restarted and a series of experiments and measurements were made. According to the data from 29 criticality experiments, the criticality element number and mass were calculated, the control rod reactivity worth were measured by period method and rod compensate method, reactivity worth of safety rod and safety block were measured using reactivity instrument; the reactivity worth of outer elements and radial distribution of elements were measured too. Based on all the measurements mentioned above, safety operation parameters for core 2 in DF-VI fast neutron criticality facility were conformed. (authors)

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)

Reactivity feedback evaluation of material relocations in the CABRI-1 experiments with fuel worth distributions from SNR-300

International Nuclear Information System (INIS)

Royl, P.; Pfrang, W.; Struwe, D.

1991-01-01

The fuel relocations from the CABRI-1 experiments with irradiated fuel that had been evaluated from the hodoscope measurements were used together with fuel reactivity worth distributions from the SNR-300 to estimate the reactivity effect which these motions would have if they occurred in SNR-300 at the same relative distance to the peak power as in CABRI. The procedure for the reactivity evaluation is outlined including the assumptions made for fuel mass conservation. The results show that the initial fuel motion yields always negative reactivities. They also document the mechanism for a temporary reactivity increase by in-pin fuel flow in some transient overpower tests. This mechanism, however, never dominates, because material accumulates always sufficiently above the peak power point. Thus, the late autocatalytic amplifications of voiding induced power excursions by compactive in-pin fuel flow, that had been simulated in bounding loss of flow analyses for SNR-300, have no basis at all when considering the results from the CABRI-1 experiments

Evaluating data worth for ground-water management under uncertainty

Science.gov (United States)

Wagner, B.J.

1999-01-01

A decision framework is presented for assessing the value of ground-water sampling within the context of ground-water management under uncertainty. The framework couples two optimization models-a chance-constrained ground-water management model and an integer-programing sampling network design model-to identify optimal pumping and sampling strategies. The methodology consists of four steps: (1) The optimal ground-water management strategy for the present level of model uncertainty is determined using the chance-constrained management model; (2) for a specified data collection budget, the monitoring network design model identifies, prior to data collection, the sampling strategy that will minimize model uncertainty; (3) the optimal ground-water management strategy is recalculated on the basis of the projected model uncertainty after sampling; and (4) the worth of the monitoring strategy is assessed by comparing the value of the sample information-i.e., the projected reduction in management costs-with the cost of data collection. Steps 2-4 are repeated for a series of data collection budgets, producing a suite of management/monitoring alternatives, from which the best alternative can be selected. A hypothetical example demonstrates the methodology's ability to identify the ground-water sampling strategy with greatest net economic benefit for ground-water management.A decision framework is presented for assessing the value of ground-water sampling within the context of ground-water management under uncertainty. The framework couples two optimization models - a chance-constrained ground-water management model and an integer-programming sampling network design model - to identify optimal pumping and sampling strategies. The methodology consists of four steps: (1) The optimal ground-water management strategy for the present level of model uncertainty is determined using the chance-constrained management model; (2) for a specified data collection budget, the monitoring

Braided reinforced composite rods for the internal reinforcement of concrete

Science.gov (United States)

Gonilho Pereira, C.; Fangueiro, R.; Jalali, S.; Araujo, M.; Marques, P.

2008-05-01

This paper reports on the development of braided reinforced composite rods as a substitute for the steel reinforcement in concrete. The research work aims at understanding the mechanical behaviour of core-reinforced braided fabrics and braided reinforced composite rods, namely concerning the influence of the braiding angle, the type of core reinforcement fibre, and preloading and postloading conditions. The core-reinforced braided fabrics were made from polyester fibres for producing braided structures, and E-glass, carbon, HT polyethylene, and sisal fibres were used for the core reinforcement. The braided reinforced composite rods were obtained by impregnating the core-reinforced braided fabric with a vinyl ester resin. The preloading of the core-reinforced braided fabrics and the postloading of the braided reinforced composite rods were performed in three and two stages, respectively. The results of tensile tests carried out on different samples of core-reinforced braided fabrics are presented and discussed. The tensile and bending properties of the braided reinforced composite rods have been evaluated, and the results obtained are presented, discussed, and compared with those of conventional materials, such as steel.

Scram characteristics of the control rods of a pressurized water reactor under seismic conditions

International Nuclear Information System (INIS)

Fujita, Katsuhisa; Shinohara, Yoshikazu; Nakatogawa, Tetsuto; Nanbu, Kiyoshi; Nomura, Tomonori.

1987-01-01

Control rod drop verification experiments of a pressurized water reactor under seismic conditions are performed to confirm the insertion function of control rods into a core. To evaluate these tests, computer simulations are performed. A fuel assembly, control rods, guide tube and other associated structures are immersed in a water tank, and shaken by four hydraulic shakers. The scram time of control rods under seismic conditions was measured, and confirmed to meet the scram function. Moreover, vibrational response characteristics of core structures and dropping behavior of control rods in consideration of collisions are calculated by using a finite difference method. The behavior of the dropping control rods and the scram time obtained by the computer simulation show a very good agreement with the verification experimental results. (author)

Seismic analysis of control and safety rod drive mechanism

International Nuclear Information System (INIS)

Meher Prasad, A.; Jaya, K.P.; Chellapandi, P.; Rajan Babu, V.; Selvaraj, T.

2003-01-01

Control rod and its driving mechanism for a Fast Breeder Reactor is to facilitate safe shutdown of the reactor in case of emergency. A theoretical study on the seismic qualification of control and safety rod driving mechanism is carried out. Earthquake excitations under Operational Basis (ORE) and Safe Shutdown condition (SSE) are considered. The time required for the control rod to reach the bottom position in order to shut down the reaction under excited condition is traced out. The maximum displaced positions and extreme stresses in various parts of the system under excitations are evaluated. The system modeled using beam elements. The connections between different parts are modeled through rigid elements. The interaction between various parts are modeled using GAP elements. (author)

Void worths in subcritical cores cooled by lead-bismuth

International Nuclear Information System (INIS)

Wallenius, Janne; Tucek, Kamil; Gudowski, Waclaw

2001-01-01

The introduction lead-bismuth coolant in accelerator driven transmutation systems (ADS) was: good neutron economy (higher source efficiency); natural circulation possible (decay heat removal); synergy with spallation target (simplified coolant management); high temperature of boiling (larger overpower margin); smaller void worths (operation at higher k-values). This paper deals with different aspects of the void worths in JAERI ADS

Division I men and women athletes do not differ on perceptions of worth.

Science.gov (United States)

Lockhart, Barbara D; Black, Nate; Vincent, William J

2012-04-01

Historically, especially prior to the increased interest in women's athletics with the passage of Title IX in 1972, there have been negative perceptions of women as athletes. If these social perceptions still hold in part today, as is indirectly suggested by unequal press coverage and less basic support for women athletes, one might predict that collegiate female athletes would rate themselves lower on self-esteem and worth than collegiate male athletes. 176 Division I male (n = 90) and female (n = 86) athletes rated their perceptions of self on the Worth Index which measures basic human worth, personal security, performance, and physical self; these are divided into intrinsic (unconditional worth) measures and behavior or performance (conditional worth) measures. There were no significant sex differences in the ratings of any aspect of perceived worth, in contrast to prior results among non-athletes. In spite of less support given to women athletes, perhaps the long-term high-intensity competition that is required to reach Division I status tends to eliminate sex differences in self-worth.

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

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.

Plant dynamics analyses of fast reactor concept: RAPID-A without any control rod

International Nuclear Information System (INIS)

Kambe, Mitsuru

1996-01-01

Plant dynamics analyses of a fast reactor concept RAPID-A without any control rod have been demonstrated in case of reactor startup and sudden change of the primary flow rate. RAIP-A concept involves Lithium Expansion Module (LEM) for inherent reactivity feedback, Lithium Injection Module (LIM) for inherent ultimate shutdown and Lithium Release Module (LRM) for automated reactor startup. LEM consists of Quick-LEM and Slow-LEM. Slow-LEM provides with moderate reactivity addition as decreasing temperature. Quick-LEM assures quick negative reactivity feedback as increasing temperature. Plant dynamics analyses revealed that reactor power is nearly proportional to the primary flow rate even if the flow rate increases suddenly. Fully automated reactor startup from the subcritical condition has been attempted by inserting reactivity at a constant rate by LRM. Allowable rate of reactivity addition has been obtained in respect to Quick-LEM reactivity worth. (author)

Intraoperative implant rod three-dimensional geometry measured by dual camera system during scoliosis surgery.

Science.gov (United States)

Salmingo, Remel Alingalan; Tadano, Shigeru; Abe, Yuichiro; Ito, Manabu

2016-05-12

Treatment for severe scoliosis is usually attained when the scoliotic spine is deformed and fixed by implant rods. Investigation of the intraoperative changes of implant rod shape in three-dimensions is necessary to understand the biomechanics of scoliosis correction, establish consensus of the treatment, and achieve the optimal outcome. The objective of this study was to measure the intraoperative three-dimensional geometry and deformation of implant rod during scoliosis corrective surgery.A pair of images was obtained intraoperatively by the dual camera system before rotation and after rotation of rods during scoliosis surgery. The three-dimensional implant rod geometry before implantation was measured directly by the surgeon and after surgery using a CT scanner. The images of rods were reconstructed in three-dimensions using quintic polynomial functions. The implant rod deformation was evaluated using the angle between the two three-dimensional tangent vectors measured at the ends of the implant rod.The implant rods at the concave side were significantly deformed during surgery. The highest rod deformation was found after the rotation of rods. The implant curvature regained after the surgical treatment.Careful intraoperative rod maneuver is important to achieve a safe clinical outcome because the intraoperative forces could be higher than the postoperative forces. Continuous scoliosis correction was observed as indicated by the regain of the implant rod curvature after surgery.

The Defect Inspection on the Irradiated Fuel Rod by Eddy Current Test

International Nuclear Information System (INIS)

Koo, D. S.; Park, Y. K.; Kim, E. K.

1996-01-01

The eddy current test(ECT) probe of differential encircling coil type was designed and fabricated, and the optimum condition of ECT was derived for the examination of the irradiated fuel rod. The correlation between ECT test frequency and phase and amplitude was derived by performing the test of the standard rig that includes inner notches, outer notches and through-holes. The defect of through-hole was predicted by ECT at the G33-N2 fuel rod irradiated in the Kori-1 nuclear power reactor. The metallographic examination on the G33-N2 fuel rod was Performed at the defect location predicted by ECT. The result of metallographic examination for the G33-N2 fuel rod was in good agreement with that of ECT. This proves that the evaluation for integrity of irradiated fuel rod by ECT is reliable

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

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

Calculational and experimental experience on core management of experimental fast reactor 'JOYO'

International Nuclear Information System (INIS)

Yoshida, A.; Arii, Y.; Shono, A.; Suzuki, S.; Kinjo, K.

1992-01-01

For the core management of JOYO Mark-II, many core characteristics have been calculated with the core management code system 'MAGI', and measurements have also been carried out at each duty operation cycle. From the evaluation of these results, the characteristics of core parameters such as criticality, reactivity coefficients, and control rod worth can be predicted accurately as followings; excess reactivity: ± 0.1% Δk/k, outlet temperature of subassembly: ±10degC, fuel burn-up: ±5%, control rod worth: ±5%. As a result, we can not only get steady operation of JOYO but also perform various irradiation tests with satisfied conditions. This paper presents experience obtained until now through twenty three duty cycle operations of Mark-II core in JOYO. (author)

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.)

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

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.

Contingent self-worth moderates the relationship between school stressors and psychological stress responses.

Science.gov (United States)

Ishizu, Kenichiro

2017-04-01

This study examined the moderating role of contingent self-worth on the relationships between school stressors and psychological stress responses among Japanese adolescents. A total of 371 Japanese junior high school students (184 boys and 187 girls, M age  = 12.79 years, SD = 0.71) completed the Japanese version of the Self-Worth Contingency Questionnaire and a mental health checklist at two points separated by a two-month interval. Hierarchical multiple regression analyses were then used to determine whether contingent self-worth moderated the relationship between school stressors and psychological stress responses. The results indicated that, when psychological stress responses were controlled for at Time 1, contingent self-worth did not predict the psychological stress responses at Time 2. However, a two-way interaction between contingent self-worth and stressors was found to significantly influence psychological stress responses, thus indicating that stressors had a stronger impact on psychological stress responses among those with high contingent self-worth compared to those with low contingent self-worth. Copyright © 2017 The Foundation for Professionals in Services for Adolescents. Published by Elsevier Ltd. All rights reserved.

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.)

Future plans for performance analysis and maintenance/inspection optimization of shutoff rods based on the case study of Bruce Power Unit-3 Shutoff Rod 5 inspection

International Nuclear Information System (INIS)

Nasimi, E.; Gabbar, H.A.

2011-01-01

Shutdown System 1 (SDS1) is a preferred method for a quick shutdown of nuclear fission process in CANDU (CANada Deuterium Uranium) reactor units. Failure of a routine SDS1 safety test during Fall 2009 outage resulted in the need to develop and execute a new methodology for Shutoff Rod inspection and re-evaluate the known degradation mechanisms and failure modes. This paper describes the development of this methodology and the obtained results. It also proposes several alternative solutions for the future performance analysis and maintenance/inspection optimization for SDS1 Shutoff Rods based on the Bruce Power Unit-3 Shutoff Rod 5 case study. (author)

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

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

Tabular method of critical heat flux description in square packing rod bundles

International Nuclear Information System (INIS)

Bobkov, V.P.; Smogalev, I.P.

2003-01-01

Elaborations of harnessing tabular method for the description and calculation of critical heat fluxes in square packing rod bundles are presented. The tabular method for fuel rod triangular assemblies derived from using basic table for critical heat fluxes in triangular fuel assemblies demonstrates good results. For the harnessing tabular method in square packing rod bundles correction functions reflecting specific geometry were found. Comparative evaluations of calculated values for the critical heat fluxes with experimental ones are presented. Good agreement of calculations with experiments is noted in all range of parameters [ru

Development of examination technique for oxide layer thickness measurement of irradiated fuel rods

International Nuclear Information System (INIS)

Koo, D. S.; Park, S. W.; Kim, J. H.; Seo, H. S.; Min, D. K.; Kim, E. K.; Chun, Y. B.; Bang, K. S.

1999-06-01

Technique for oxide layer thickness measurement of irradiated fuel rods was developed to measure oxide layer thickness and study characteristic of fuel rods. Oxide layer thickness of irradiated fuels were measured, analyzed. Outer oxide layer thickness of 3 cycle-irradiated fuel rods were 20 - 30 μm, inner oxide layer thickness 0 - 10 μm and inner oxide layer thickness on cracked cladding about 30 μm. Oxide layer thickness of 4 cycle-irradiated fuel rods were about 2 times as thick as those of 1 cycle-irradiated fuel rods. Oxide layer on lower region of irradiated fuel rods was thin and oxide layer from lower region to upper region indicated gradual increase in thickness. Oxide layer thickness from 2500 to 3000 mm showed maximum and oxide layer thickness from 3000 to top region of irradiated fuel rods showed decreasing trend. Inner oxide layer thicknesses of 4 cycle-irradiated fuel rod were about 8 μm at 750 - 3500 mm from the bottom end of fuel rod. Outer oxide layer thickness were about 8 μm at 750 - 1000 mm from the bottom end of fuel rod. These indicated gradual increase up to upper region from the bottom end of fuel rod. These indicated gradual increase up to upper region from the bottom end of fuel. Oxide layer thickness technique will apply safety evaluation and study of reactor fuels. (author). 6 refs., 14 figs

Rod internal pressure quantification and distribution analysis using Frapcon

Energy Technology Data Exchange (ETDEWEB)

Jessee, Matthew Anderson [ORNL; Wieselquist, William A [ORNL; Ivanov, Kostadin [Pennsylvania State University, University Park

2015-09-01

This report documents work performed supporting the Department of Energy (DOE) Office of Nuclear Energy (NE) Fuel Cycle Technologies Used Fuel Disposition Campaign (UFDC) under work breakdown structure element 1.02.08.10, ST Analysis. In particular, this report fulfills the M4 milestone M4FT- 15OR0810036, Quantify effects of power uncertainty on fuel assembly characteristics, within work package FT-15OR081003 ST Analysis-ORNL. This research was also supported by the Consortium for Advanced Simulation of Light Water Reactors (http://www.casl.gov), an Energy Innovation Hub (http://www.energy.gov/hubs) for Modeling and Simulation of Nuclear Reactors under U.S. Department of Energy Contract No. DE-AC05-00OR22725. The discharge rod internal pressure (RIP) and cladding hoop stress (CHS) distributions are quantified for Watts Bar Nuclear Unit 1 (WBN1) fuel rods by modeling core cycle design data, operation data (including modeling significant trips and downpowers), and as-built fuel enrichments and densities of each fuel rod in FRAPCON-3.5. A methodology is developed which tracks inter-cycle assembly movements and assembly batch fabrication information to build individual FRAPCON inputs for each evaluated WBN1 fuel rod. An alternate model for the amount of helium released from the zirconium diboride (ZrB2) integral fuel burnable absorber (IFBA) layer is derived and applied to FRAPCON output data to quantify the RIP and CHS for these types of fuel rods. SCALE/Polaris is used to quantify fuel rodspecific spectral quantities and the amount of gaseous fission products produced in the fuel for use in FRAPCON inputs. Fuel rods with ZrB2 IFBA layers (i.e., IFBA rods) are determined to have RIP predictions that are elevated when compared to fuel rod without IFBA layers (i.e., standard rods) despite the fact that IFBA rods often have reduced fill pressures and annular fuel pellets. The primary contributor to elevated RIP predictions at burnups less than and greater than 30 GWd

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

International Nuclear Information System (INIS)

Leech, W.J.; Kaiser, R.S.

1980-01-01

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

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

Effect of fission fragment on thermal conductivity via electrons with an energy about 0.5 MeV in fuel rod gap

Directory of Open Access Journals (Sweden)

F Golian

2017-02-01

Full Text Available The heat transfer process from pellet to coolant is one of the important issues in nuclear reactor. In this regard, the fuel to clad gap and its physical and chemical properties are effective factors on heat transfer in nuclear fuel rod discussion. So, the energy distribution function of electrons with an energy about 0.5 MeV in fuel rod gap in Busherhr’s VVER-1000 nuclear reactor was investigated in this paper. Also, the effect of fission fragments such as Krypton, Bromine, Xenon, Rubidium and Cesium on the electron energy distribution function as well as the heat conduction via electrons in the fuel rod gap have been studied. For this purpose, the Fokker- Planck equation governing the stochastic behavior of electrons in absorbing gap element has been applied in order to obtain the energy distribution function of electrons. This equation was solved via Runge-Kutta numerical method. On the other hand, the electron energy distribution function was determined by using Monte Carlo GEANT4 code. It was concluded that these fission fragments have virtually insignificant effect on energy distribution of electrons and therefore, on thermal conductivity via electrons in the fuel to clad gap. It is worth noting that this result is consistent with the results of other experiments. Also, it is shown that electron relaxation in gap leads to decrease in thermal conductivity via electrons

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...

Photonic crystal rod fibers: Understanding a new class of active optical waveguides

DEFF Research Database (Denmark)

Laurila, Marko

core modes and the cladding band as the underlying mechanics to ensure SM operation of the new rod fiber design (85 μm core diameter), which was developed during this thesis work. The second focus of this work is the study of the new ytterbiumdoped rod fiber design under active operation. Performance...... of the rod fiber is evaluated in high power laser and laser amplifier configurations. The high power rod amplifier setup including the seed source is developed and characterized. Results obtained from the rod fiber showed simultaneously SM, near diffraction limited output beam quality with high average power...... and pulse energy generation using both laser and laser amplifier configurations. Modal instabilities (MIs) in high power fiber amplifiers are discussed, and a memory effect of the MI threshold level together with a recovery method and evidence of improved performance while suppressing MIs are reported...

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

Evaluation of the professional worth of scientific papers, their citation responding and the publication authority

Czech Academy of Sciences Publication Activity Database

Šesták, Jaroslav; Fiala, J.; Gavrichev, K.S.

2018-01-01

Roč. 131, č. 1 (2018), s. 463-471 ISSN 1388-6150 Institutional research plan: CEZ:AV0Z10100502 Institutional support: RVO:68378271 Keywords : professional worth * citation response * quantity versus quality * impact factors * predatory journals * databases * JTAC * fields venture Subject RIV: BM - Solid Matter Physics ; Magnetism OBOR OECD: Condensed matter physics (including formerly solid state physics, supercond.) Impact factor: 1.953, year: 2016

Transgenic Expression of Constitutively Active RAC1 Disrupts Mouse Rod Morphogenesis

Science.gov (United States)

Song, Hongman; Bush, Ronald A.; Vijayasarathy, Camasamudram; Fariss, Robert N.; Kjellstrom, Sten; Sieving, Paul A.

2014-01-01

Purpose. Dominant-active RAC1 rescues photoreceptor structure in Drosophila rhodopsin-null mutants, indicating an important role in morphogenesis. This report assesses the morphogenetic effect of activated RAC1 during mammalian rod photoreceptor development using transgenic mice that express constitutively active (CA) RAC1. Methods. Transgenic mice were generated by expressing CA RAC1 under control of the Rhodopsin promoter, and morphological features of the photoreceptors were evaluated by histology, immunohistochemistry, and transmission electron microscopy. Function was evaluated by electroretinography. Potential protein partners of CA RAC1 were identified by co-immunoprecipitation of retinal extracts. Results. Constitutively active RAC1 expression in differentiating rods disrupted outer retinal lamination as early as postnatal day (P)6, and many photoreceptor cell nuclei were displaced apically into the presumptive subretinal space. These photoreceptors did not develop normal inner and outer segments and had abnormal placement of synaptic elements. Some photoreceptor nuclei were also mislocalized into the inner nuclear layer. Extensive photoreceptor degeneration was subsequently observed in the adult animal. Constitutively active RAC1 formed a complex with the polarity protein PAR6 and with microtubule motor dynein in mouse retina. The normal localization of the PAR6 complex was disrupted in CA RAC1-expressing rod photoreceptors. Conclusions. Constitutively active RAC1 had a profound negative effect on mouse rod cell viability and development. Rod photoreceptors in the CA RAC1 retina exhibited a defect in polarity and migration. Constitutively active RAC1 disrupted rod morphogenesis and gave a phenotype resembling that found in the Crumbs mutant. PAR6 and dynein are two potential downstream effectors that may be involved in CA RAC1-mediated defective mouse photoreceptor morphogenesis. PMID:24651551

DANCOFF-3, Dancoff Correction for Cylindrical Fuel Rod at H2O Gaps and for Fuel Clusters

International Nuclear Information System (INIS)

Sauer, A.

1989-01-01

1 - Nature of physical problem solved: Calculation of the Dancoff correction for cylindrical fuel rods in square and hexagonal infinite lattices, for fuel element rods near water gaps, and for fuel rod clusters. 2 - Method of solution: Evaluation by direct numerical integration over the moderator region. 3 - Restrictions on the complexity of the problem: For every rod arrangement at most 100 cases with different materials cross- sections

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

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.)

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

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

Reliability Worth Analysis of Distribution Systems Using Cascade Correlation Neural Networks

DEFF Research Database (Denmark)

Heidari, Alireza; Agelidis, Vassilios; Pou, Josep

2018-01-01

Reliability worth analysis is of great importance in the area of distribution network planning and operation. The reliability worth's precision can be affected greatly by the customer interruption cost model used. The choice of the cost models can change system and load point reliability indices....... In this study, a cascade correlation neural network is adopted to further develop two cost models comprising a probabilistic distribution model and an average or aggregate model. A contingency-based analytical technique is adopted to conduct the reliability worth analysis. Furthermore, the possible effects...

Core concepts for ''zero-sodium-void-worth core'' in metal fuelled fast reactor

International Nuclear Information System (INIS)

Chang, Y.I.; Hill, R.N.; Fujita, E.K.; Wade, D.C.; Kumaoka, Y.; Suzuki, M.; Kawashima, M.; Nakagawa, H.

1991-01-01

Core design options to reduce the sodium void worth in metal fueled LMRs are investigated. Two core designs which achieve a zero sodium void worth are analyzed in detail. The first design is a ''pancaked'' and annular core with enhanced transuranic burning capabilities; the high leakage in this design yields a low breeding ratio and small void worth. The second design is an axially multilayered annular core which is fissile self-sufficient; in this design, the upper and lower core regions are neutronically decoupled for reduced void worth while fissile self-sufficiency is achieved using internal axial blankets plus external radial and axial blanket zones. The neutronic performance characteristics of these low void worth designs are assessed here; their passive safety properties are discussed in a companion paper. 16 refs., 2 figs., 3 tabs

Core concepts for 'zero-sodium-void-worth core' in metal fuelled fast reactor

International Nuclear Information System (INIS)

Chang, Y.I.; Hill, R.N.; Fujita, E.K.; Wade, D.C.; Kumaoka, Y.; Suzuki, M.; Kawashima, M.; Nakagawa, H.

1991-01-01

Core design options to reduce the sodium void worth in metal fuelled LMRs are investigated. Two core designs which achieve a zero sodium void worth are analyzed in detail. The first design is a 'pancaked' and annular core with enhanced transuranic burning capabilities; the high leakage in this design yields a low breeding ratio and small void worth. The second design is an axially multilayered annular core which is fissile self-sufficient; in this design, the upper and lower core regions are neutronically decoupled for reduced void worth while fissile self-sufficiency is achieved using internal axial blankets plus external radial and axial blanket-zones. The neutronic performance characteristics of these low void worth designs are assessed here; their passive safety properties are discussed in a companion paper. (author)

Non-parametric order statistics method applied to uncertainty propagation in fuel rod calculations

International Nuclear Information System (INIS)

Arimescu, V.E.; Heins, L.

2001-01-01

Advances in modeling fuel rod behavior and accumulations of adequate experimental data have made possible the introduction of quantitative methods to estimate the uncertainty of predictions made with best-estimate fuel rod codes. The uncertainty range of the input variables is characterized by a truncated distribution which is typically a normal, lognormal, or uniform distribution. While the distribution for fabrication parameters is defined to cover the design or fabrication tolerances, the distribution of modeling parameters is inferred from the experimental database consisting of separate effects tests and global tests. The final step of the methodology uses a Monte Carlo type of random sampling of all relevant input variables and performs best-estimate code calculations to propagate these uncertainties in order to evaluate the uncertainty range of outputs of interest for design analysis, such as internal rod pressure and fuel centerline temperature. The statistical method underlying this Monte Carlo sampling is non-parametric order statistics, which is perfectly suited to evaluate quantiles of populations with unknown distribution. The application of this method is straightforward in the case of one single fuel rod, when a 95/95 statement is applicable: 'with a probability of 95% and confidence level of 95% the values of output of interest are below a certain value'. Therefore, the 0.95-quantile is estimated for the distribution of all possible values of one fuel rod with a statistical confidence of 95%. On the other hand, a more elaborate procedure is required if all the fuel rods in the core are being analyzed. In this case, the aim is to evaluate the following global statement: with 95% confidence level, the expected number of fuel rods which are not exceeding a certain value is all the fuel rods in the core except only a few fuel rods. In both cases, the thresholds determined by the analysis should be below the safety acceptable design limit. An indirect

Prototypical Rod Consolidation Demonstration Project

International Nuclear Information System (INIS)

1993-05-01

The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 1 discusses the following topics: the background of the project; test program description; summary of tests and test results; problem evaluation; functional requirements confirmation; recommendations; and completed test documentation for tests performed in Phase 3

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)

New models of droplet deposition and entrainment for prediction of CHF in cylindrical rod bundles

Energy Technology Data Exchange (ETDEWEB)

Zhang, Haibin, E-mail: hb-zhang@xjtu.edu.cn [School of Chemical Engineering and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Department of Chemical Engineering, Imperial College, London SW7 2BY (United Kingdom); Hewitt, G.F. [Department of Chemical Engineering, Imperial College, London SW7 2BY (United Kingdom)

2016-08-15

Highlights: • New models of droplet deposition and entrainment in rod bundles is developed. • A new phenomenological model to predict the CHF in rod bundles is described. • The present model is well able to predict CHF in rod bundles. - Abstract: In this paper, we present a new set of model of droplet deposition and entrainment in cylindrical rod bundles based on the previously proposed model for annuli (effectively a “one-rod” bundle) (2016a). These models make it possible to evaluate the differences of the rates of droplet deposition and entrainment for the respective rods and for the outer tube by taking into account the geometrical characteristics of the rod bundles. Using these models, a phenomenological model to predict the CHF (critical heat flux) for upward annular flow in vertical rod bundles is described. The performance of the model is tested against the experimental data of Becker et al. (1964) for CHF in 3-rod and 7-rod bundles. These data include tests in which only the rods were heated and data for simultaneous uniform and non-uniform heating of the rods and the outer tube. It was shown that the predicted CHFs by the present model agree well with the experimental data and with the experimental observation that dryout occurred first on the outer rods in 7-rod bundles. It is expected that the methodology used will be generally applicable in the prediction of CHF in rod bundles.

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

The Economics of Comparable Worth.

Science.gov (United States)

Killingsworth, Mark R.

This document concludes that the basic difficulty with comparable worth is that it is an ill-conceived solution to a serious problem and that alternative policies, such as equal employment opportunity legislation or application of antitrust laws, provide means of addressing employment discrimination that are both more effective and less likely to…

Rod displacement measurements by x-ray CT and its impact on thermal-hydraulics in tight-lattice rod bundle (Joint research)

International Nuclear Information System (INIS)

Mitsutake, Toru; Misawa, Takeharu; Kureta, Masatoshi; Akimoto, Hajime

2005-06-01

In tight-lattice simulated rod bundles with about 1 mm gap between rods, a rod displacement might affect thermal-hydraulic characteristics since the displacement has a strong impact on the flow area change along the heated section. It should be important to estimate how large the rod position displacement could quantitatively affect critical power for the tight-lattice rod bundle from the point of improvement of prediction capability of subchannel analysis. In the present study, the inside-structure observation of the simulated seven-rod bundle of Reduced Moderation Water Reactor (RMWR) was made through the whole length of the test assembly. Based on the measured rod position data, the relation between the rod position displacement and the heat transfer characteristics was investigated experimentally and through the two kinds of subchannel analysis, the nominal rod position case and the measured rod position case, the effect on the predicted critical power was estimated. The high-energy X-ray computer tomograph (CT) of Fuels Monitoring Facilities (FMF) at the O-arai Engineering Center in Japan Nuclear Cycle Institute (JNC) was applied for the inside-structure observation of the test assembly. The CT view of the cross sections within the test assembly assured the hexagonal rod position arrangement was almost the same as expected by design. The measured data with the X-ray CT facility showed that all rod displacements were small, 0.5 millimeters at maximum and 0.2 millimeters in average. In the heat transfer experiments for the seven-rod bundle, the boiling transition (BT) position and the rod surface temperature behavior was measured. All thermocouples on the center rod downstream from the BT-onset axial height showed almost simultaneous temperature increase due to BT. And the thermocouples located on the same axial heights showed quite similar time-variation behaviors in the vapor cooling heat transfer regime. These results demonstrated the effect of the

Evaluation of CHF experimental data for non-square lattice 7-rod bundles

Energy Technology Data Exchange (ETDEWEB)

Hwang, Dae Hyun; Yoo, Y. J.; Kim, K. K.; Zee, S. Q

2001-01-01

A series of CHF experiments are conducted for 7-rod hexagonal test bundles in order to investigate the CHF characteristics of self-sustained square finned (SSF) rod bundles. The experiments are performed in the freon-loop and water-loop located at IPPE in Russia, and 609 data of freon-12 and 229 data of water are obtained from 7 kinds of test bundles classified by the combination of heated length and axial/radial power distributions. As the result of the evaluation of four representative CHF correlations, the EPRI-1 correlation reveals good prediction capability for SSF test bundles. The inlet parameter CHF correlation suggested by IPPE calculates the mean and the standard deviation of P/M for uniformly heated test bundles as 1.002 and 0.049, respectively. In spite of its excellent accuracy, the correlation has a discontinuity point at the boundary between the low velocity and high velocity conditions. KAERI's inlet parameter correlation eliminates this defect by introducing the complete evaporation model at low velocity condition, and calculates the mean and standard deviation of P/M as 0.095 and 0.062 for uniformly heated 496 data points, respectively. The mean/standard deviation of local parameter CHF correlations suggested by IPPE and KAERI are evaluated as 1.023/0.178 and 1.002/0.158, respectively. The inlet parameter correlation developed from uniformly heated test bundles tends to under-predict CHF about 3% for axially non-uniformly heated test bundles. On the other hand, the local parameter correlation reveals large scattering of P/M, and requires re-optimization of the correlation for non-uniform axial power distributions. As the result of the analysis of experimental data, it reveals that the correction model of axial power shapes suggested by IPPE is applicable to the inlet parameter correlations. For the test bundle of radial non-uniform power distribution, the physically unexpected results are obtained at some experimental conditions. In addition

Control rod ejection analysis during a depressurization accident and the development of a rod-ejection-preventing device

International Nuclear Information System (INIS)

Mitake, S.; Itoh, K.; Fukushima, H.; Inoue, T.

1982-01-01

The control rods used for the experimental VHTR are suspended in the core by means of flexible steel cables and it is conceivable that an accidental rod ejection could occur due to a depressurization accident. The computer code AFLADE was developed in order to analyze the possibility of accidental rod ejection, and several studies were performed. The parametric study results showed that the adopted design condition for the VHTR core will not cause a rod ejection accident. In parallel with these accident analyses, a rod-ejection-preventing device was developed in preparation for a hypothetical accident, and its function was verified by the component tests

Internal hydriding in irradiated defected Zircaloy fuel rods: A review (LWBR Development Program)

International Nuclear Information System (INIS)

Clayton, J.C.

1987-10-01

Although not a problem in recent commercial power reactors, including the Shippingport Light Water Breeder Reactor, internal hydriding of Zircaloy cladding was a persistent cause of gross cladding failures during the 1960s. It occurred in the fuel rods of water-cooled nuclear power reactors that had a small cladding defect. This report summarizes the experimental findings, causes, mechanisms, and methods of minimizing internal hydriding in defected Zircaloy-clad fuel rods. Irradiation test data on the different types of defected fuel rods, intentionally fabricated defected and in-pile operationally defected rods, are compared. Significant factors affecting internal hydriding in defected Zircaloy-clad fuel rods (defect hole size, internal and external sources of hydrogen, Zircaloy cladding surface properties, nickel alloy contamination of Zircaloy, the effect of heat flux and fluence) are discussed. Pertinent in-pile and out-of-pile test results from Bettis and other laboratories are used as a data base in constructing a qualitative model which explains hydrogen generation and distribution in Zircaloy cladding of defected water-cooled reactor fuel rods. Techniques for minimizing internal hydride failures in Zircaloy-clad fuel rods are evaluated

Self-worth, perceived competence, and behaviour problems in children with cerebral palsy

OpenAIRE

Schuengel, C.; Voorman, J.; Stolk, J.; Dallmeijer, A.J.; Vermeer, A; Becher, J.

2006-01-01

Purpose. To examine the relevance of physical disabilities for self-worth and perceived competence in children with cerebral palsy (CP), and to examine associations between behaviour problems and self-worth and perceived competence. Methods. The Harter scales for self-worth and perceived competence and a new scale for perceived motor competence were used in a sample of 80 children with CP. Their motor functioning was assessed with the Gross Motor Functioning Measure (GMFM) and behaviour probl...

The Relationship of Coping, Self-Worth, and Subjective Well-Being: A Structural Equation Model

Science.gov (United States)

Smedema, Susan Miller; Catalano, Denise; Ebener, Deborah J.

2010-01-01

The purpose of this study was to determine the relationship between various coping-related variables and the evaluation of self-worth and subjective well-being among persons with spinal cord injury. Positive coping variables included hope, proactive coping style, and sense of humor, whereas negative coping variables included perceptions of stress,…

Single-phase convective heat transfer in rod bundles

International Nuclear Information System (INIS)

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

2008-01-01

The convective heat transfer for turbulent flow through rod bundles representative of nuclear fuel rods used in pressurized water reactors is examined. The rod bundles 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, which create swirling flow in the rod bundle, as well as disc and standard support grids are investigated. Single-phase convective heat transfer coefficients are measured for flow downstream of support grids in a rod bundle. The rods are heated using direct resistance heating, and a bulk axial flow of air is used to cool the rods in the rod bundle. Air is used as the working fluid instead of water to reduce the power required to heat the rod bundle. Results indicate heat transfer enhancement for up to 10 hydraulic diameters downstream of the support grids. A general correlation is developed to predict the heat transfer development downstream of support grids. In addition, circumferential variations in heat transfer coefficients result in hot streaks that develop on the rods downstream of split-vane pair support grids

Single-phase convective heat transfer in rod bundles

Energy Technology Data Exchange (ETDEWEB)

Holloway, Mary V. [Mechanical Engineering Department, United States Naval Academy, 590 Holloway Rd., Annapolis, MD 21402 (United States)], E-mail: holloway@usna.edu; Beasley, Donald E. [Mechanical Engineering Department, Clemson University, Clemson, SC 29634 (United States); Conner, Michael E. [Westinghouse Nuclear Fuel, 5801 Bluff Road, Columbia, SC 29250 (United States)

2008-04-15

The convective heat transfer for turbulent flow through rod bundles representative of nuclear fuel rods used in pressurized water reactors is examined. The rod bundles 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, which create swirling flow in the rod bundle, as well as disc and standard support grids are investigated. Single-phase convective heat transfer coefficients are measured for flow downstream of support grids in a rod bundle. The rods are heated using direct resistance heating, and a bulk axial flow of air is used to cool the rods in the rod bundle. Air is used as the working fluid instead of water to reduce the power required to heat the rod bundle. Results indicate heat transfer enhancement for up to 10 hydraulic diameters downstream of the support grids. A general correlation is developed to predict the heat transfer development downstream of support grids. In addition, circumferential variations in heat transfer coefficients result in hot streaks that develop on the rods downstream of split-vane pair support grids.

Reactivity worth measurement of the control blades of the University of Florida training reactor

International Nuclear Information System (INIS)

Quintero-Leyva, Barbaro

1997-01-01

A series of experiments were carried out in order to measure the reactivity worth of the safety and regulating blades of the University of Florida Training Reactor (UFTR) using the Inverse Kinetics, the Inverse Kinetics-Rod Drop method and the Power Ratio. The reactor's own instrumentation (compensated ion chamber) and an independent counting system (fission chamber) were used. A very smooth exponential decay of the flux was observed after 6s of the beginning of the transients using the reading of the reactor detector. The results of the measurements of the reactivity using both detectors were consistent and in good agreement. The compensated ion chamber showed a very smooth exponential behavior; this suggests that if we could record the power for a small sample time, say 0.1 s from the beginning of the transient, several additional research projects could be accomplished. First, precise intercomparison of the methods could be achieved if the statistics level is acceptable. Second, a precise description of the bouncing of the blades and its effects on the reactivity could be achieved. Finally, the design of a reactivity-meter could be based on such study. (author)

Fabrication of internally instrumented reactor fuel rods

International Nuclear Information System (INIS)

Schmutz, J.D.; Meservey, R.H.

1975-01-01

Procedures are outlined for fabricating internally instrumented reactor fuel rods while maintaining the original quality assurance level of the rods. Instrumented fuel rods described contain fuel centerline thermocouples, ultrasonic thermometers, and pressure tubes for internal rod gas pressure measurements. Descriptions of the thermocouples and ultrasonic thermometers are also contained

Vortex Noise from Rotating Cylindrical Rods

Science.gov (United States)

Stowell, E Z; Deming, A F

1935-01-01

A series of round rods of the some diameter were rotated individually about the mid-point of each rod. Vortices are shed from the rods when in motion, giving rise to the emission of sound. With the rotating system placed in the open air, the distribution of sound in space, the acoustical power output, and the spectral distribution have been studied. The frequency of emission of vortices from any point on the rod is given by the formula von Karman. From the spectrum estimates are made of the distribution of acoustical power along the rod, the amount of air concerned in sound production, the "equivalent size" of the vortices, and the acoustical energy content for each vortex.

Damage and failure of unirradiated and irradiated fuel rods tested under film boiling conditions

International Nuclear Information System (INIS)

Mehner, A.S.; Hobbins, R.R.; Seiffert, S.L.; MacDonald, P.E.; McCardell, R.K.

1979-01-01

Power-cooling-mismatch experiments are being conducted as part of the Thermal Fuels Behavior Program in the Power Burst Facility at the Idaho National Engineering Laboratory to evaluate the behavior of unirradiated and previously irradiated light water reactor fuel rods tested under stable film boiling conditions. The observed damage that occurs to the fuel rod cladding and the fuel as a result of film boiling operation is reported. Analyses performed as a part of the study on the effects of operating failed fuel rods in film boiling, and rod failure mechanisms due to cladding embrittlement and cladding melting upon being contacted by molten fuel are summarized

Vibrational characteristics and wear of fuel rods

International Nuclear Information System (INIS)

Schmugar, K.L.

1977-01-01

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

The improvement of control rod in experimental fast reactor JOYO. The development of a sodium bonded type control rod

Energy Technology Data Exchange (ETDEWEB)

Soga, T.; Miyakawa, S.; Mitsugi, T. [Japan Nuclear Cycle Development Inst., Oarai Engineering Center, Irradiation Center, Irradiation and Administration Section, Oarai, Ibaraki (Japan)

1999-06-01

that the cladding strength sufficient to withstand stress accounting for decreased thickness by the ACCI zone. (5) The wet wash and storage method was selected for disposing of the spent sodium bonded control rods, based upon experimental results at the JOYO facilities. The effects from storing sodium bonded control rods in wet storage were evaluated. The results indicated that these effect would not pose a safety problem. (author)

Drive-in device for long thin rods into narrow cavitations, especially for control-shutdown rods e.g. of nuclear reactors

International Nuclear Information System (INIS)

Flessner, H.; Paeserack, U.

1974-01-01

The auxiliary device serves as holder for long and thin rods, e.g. control rods, transported hanging in bundles, when these are lowered into narrow cavities. It is constructed as a rod grab vertically movable at the end of a guide tube. A comb-shaped trap in connection with a guide rod serves for lateral support of the lower ends of the rods hanging on the grab. This guide rod can be moved in vertical direction by means of two pairs of convex rollers resting on the inner guide tube. In addition, the guide rod has a prolongation carrying a traverse by means of an abutment on the lower end. With these auxiliaries amongst others, the trap can be brought into a horizontal position by turning around an axis with the control rods meshing with the teeth of the trap while the parallelism of the rods is kept up during transport. (DG) [de

A longitudinal assessment of the links between physical activity and physical self-worth in adolescent females.

Science.gov (United States)

Raudsepp, Lennart; Neissaar, Inga; Kull, Merike

2013-01-01

A longitudinal framework was used to examine the hypotheses of (1) whether physical activity predicts changes in physical self-worth or (2) whether physical self-worth predicts changes in physical activity in adolescent girls. Participants (n=272) completed measures of physical self-worth and participation in physical activities at three different points spanning a two-year interval. A cross-lagged panel model using structural equation modelling analyses indicated that physical self-worth predicted subsequent physical activity and physical activity in turn predicted subsequent physical self-worth across time. Findings demonstrated a reciprocal relationship between physical self-worth and physical activity during early adolescence. This study supports the use of the reciprocal effects model (REM) in gaining an understanding of the cross-lagged relationships between physical self-worth and participation in physical activities amongst adolescent girls.

Control rod excess withdrawal prevention device

International Nuclear Information System (INIS)

Takayama, Yoshihito.

1992-01-01

Excess withdrawal of a control rod of a BWR type reactor is prevented. That is, the device comprises (1) a speed detector for detecting the driving speed of a control rod, (2) a judging circuit for outputting an abnormal signal if the driving speed is greater than a predetermined level and (3) a direction control valve compulsory closing circuit for controlling the driving direction of inserting and withdrawing a control rod based on an abnormal signal. With such a constitution, when the with drawing speed of a control rod is greater than a predetermined level, it is detected by the speed detector and the judging circuit. Then, all of the direction control valve are closed by way of the direction control valve compulsory closing circuit. As a result, the operation of the control rod is stopped compulsorily and the withdrawing speed of the control rod can be lowered to a speed corresponding to that upon gravitational withdrawal. Accordingly, excess withdrawal can be prevented. (I.S)

Hollow rods for the oil producing industry

Energy Technology Data Exchange (ETDEWEB)

Khalimova, L M; Elyasheva, M A

1970-01-01

Hollow sucker rods have several advantages over conventional ones. The hollow rods actuate the well pump and at the same time conduct produced fluids to surface. When paraffin deposition occurs, it can be minimized by injecting steam, hot oil or hot water into the hollow rod. Other chemicals, such as demulsifiers, scale inhibitors, corrosion inhibitors, etc., can also be placed in the well through the hollow rods. This reduces cost of preventive treatments, reduces number of workovers, increases oil production, and reduces cost of oil. Because the internal area of the rod is small, the passing liquids have a high velocity and thereby carry sand and dirt out of the well. This reduces pump wear between the piston and the plunger. Specifications of hollow rods, their operating characteristics, and results obtained with such rods under various circumstances are described.

ELECTRIC FIELD MEASUREMENT IN ROD-DISCONTINUED ...

African Journals Online (AJOL)

2014-06-30

Jun 30, 2014 ... the electrogeometrical model using a laboratory experimental rod-plane air gap arrangement with a lightning conductor (Franklin rod or horizontal conductor). The stepped leader could be represented by the rod electrode under a negative lightning impulse voltage having a level leading to breakdown with ...

Analysis of JUPITER experiment in ZPPR-9

Energy Technology Data Exchange (ETDEWEB)

NONE

1980-09-15

Information and data from the ZPPR-9 reactor JUPITER experiment are presented concerning a general description of data and methods; criticality; reaction rate ratio and reaction rate distribution; Doppler and sample reactivity worth; sodium void worth; and control rod worth.

Method of inspecting control rod drive mechanism

International Nuclear Information System (INIS)

Sato, Tomomi; Tatemichi, Shin-ichiro; Hasegawa, Hidenobu.

1988-01-01

Purpose: To conduct inspection for control rod drives and fuel handling operations in parallel without taking out the entire fuel, while maintaining the reactor in a subcritical state. Method: Control rod drives are inspected through the release of connection between control rods and control rod drives, detachment and dismantling of control rod drives, etc. In this case, structural materials having neutron absorbing power equal to or greater than the control rods are inserted into the gap after taking out fuels. Since the structural materials have neutron absorbing portion, subcriticality is maintained by the neutron absorbing effect. Accordingly, there is no requirement for taking out all of the fuels, thereby enabling to check the control rod drives and conduct handling for the fuels in parallel. As a result, the number of days required for the inspection can be shortened and it is possible to improve the working efficiency for the decomposition, inspection, etc. of the control rod drives and, thus, improve the operation efficiency of the nuclear power plant thereby attaining the predetermined purpose. (Kawakami, Y.)

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

Control rod

International Nuclear Information System (INIS)

Takeda, Toshikazu; Inoue, Kotaro.

1979-01-01

Purpose: To flatten the power distribution in the reactor core without impairing neutron economy by disposing pins containing elements of lower atomic number in the central region of a shroud and loading pins containing depleted uranium in the periphery region thereof. Constitution: The shroud has a layer of pins containing depleted uranium in the peripheral region and a layer of pins containing elements of lower atomic number such as beryllium in the central region. Heat removal from those pins containing depleted uranium and elements of lower atomic number (neutron moderator) is effected by sodium flow outside of the cladding material. The control rod operation is conducted by inserting or extracting the central portion (pins containing elements of lower atomic number such as beryllium) inside of the stainless pipe. Upon extraction of the control rod, the moderator in the central region is removed whereby high speed neutrons are no more deccelerated and the absorption rate to the depleted uranium is decreased. This can flatten the power distribution in the reactore core with the disposition of a plurality of control rods at a better neutron economy as compared with the use of neutron absorber such as boron. (Seki, T.)

Control rod

International Nuclear Information System (INIS)

Fukumoto, Takashi; Hirakawa, Hiromasa; Kawashima, Norio; Goto, Yasuyuki.

1994-01-01

Neutron absorbers are contained in a tubular member comprising, integrally a tubular portion and four corners disposed at the outer circumference of the tubular portion at every 90deg, to provide a neutron absorbing tube. A plurality of neutron absorbing tubes are arranged in parallel in the lateral direction, and adjacent corners are joined, into a blade to constitute a control rod. Such a control rod has a great structural strength, simple in the structure and relatively light in weight and can contain a great amount of neutron absorbers. Upon formation of the control rod by arranging the blades in a cross-like shape, at least a portion thereof is constituted with short neutron absorbing tubes shorter than the entire length of the blade, and gaps are formed at positions in adjacent in the axial direction. With such a constitution, there is no worry that a wing end of the blade collides against or be abraded with a fuel channel box or a fuel support. Even if fuel channels are vibrated upon scram of the reactor, such as occurrence of earthquakes, it can be inserted to the reactor easily. (N.H.)

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

Experimental determination of temperature fields in sodium-cooled rod bundles with hexagonal rod arrangement and grid spacers

International Nuclear Information System (INIS)

Moeller, R.; Tschoeke, H.; Kolodziej, M.

1977-01-01

Three-dimensional temperature fields in the claddings of sodium cooled rods were determined experimentally under representative nominal operating conditions for a SNR typical 19-rod bundle model provided with spark-eroded spacers. These experiments are required to verify thermohydraulic computer programs which will provide the output data for strength calculations of the high loaded cladding tubes. In this work the essentials are reported of the measured circumferential distributions of wall temperatures of peripheral rods. In addition the sub-channel temperatures measured over the bundle cross section are indicated, they are required to sustain codes for the global thermohydraulic design of core elements. The most important results are: 1) The whole fuel element is located within the thermal entrance length. 2) High azimuthal temperature differences were measured in the claddings of peripheral rods, which are strongly influenced by the distance between the rod and the shroud, especially for the corner rod. 3) With decreasing Pe-number ( [de

Control rod housing alignment

International Nuclear Information System (INIS)

Dixon, R.C.; Deaver, G.A.; Punches, J.R.; Singleton, G.E.; Erbes, J.G.; Offer, H.P.

1990-01-01

This patent describes a process for measuring the vertical alignment between a hole in a core plate and the top of a corresponding control rod drive housing within a boiling water reactor. It comprises: providing an alignment apparatus. The alignment apparatus including a lower end for fitting to the top of the control rod drive housing; an upper end for fitting to the aperture in the core plate, and a leveling means attached to the alignment apparatus to read out the difference in angularity with respect to gravity, and alignment pin registering means for registering to the alignment pin on the core plate; lowering the alignment device on a depending support through a lattice position in the top guide through the hole in the core plate down into registered contact with the top of the control rod drive housing; registering the upper end to the sides of the hole in the core plate; registering the alignment pin registering means to an alignment pin on the core plate to impart to the alignment device the required angularity; and reading out the angle of the control rod drive housing with respect to the hole in the core plate through the leveling devices whereby the angularity of the top of the control rod drive housing with respect to the hole in the core plate can be determined

Temperature and Stresses Estimation in Reactivity Control Rods for CAREM-25 Reactor

International Nuclear Information System (INIS)

Markiewicz, Mario; Estevez, Esteban

2000-01-01

The reactivity control rods are a critical component regarding safety.Its correct operation when required must be ensured.For this purpose, this component must maintain its operating capacity during all its residence time and under any foreseen operation condition.To evaluate the behaviour of reactivity control rods, it is necessary to analyse the demands they are exposed to, determining from the mechanical point of view, the residence time in the reactor core.In this report, using analytical calculations, the parameters affecting the performance of the reactivity control rods are analysed, with the objective of determine from the mechanical point of view, its behaviour and residence time

The Third ATLAS ROD Workshop

CERN Multimedia

Poggioli, L.

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

Temperature actuated automatic safety rod release

Science.gov (United States)

Hutter, E.; Pardini, J.A.; Walker, D.E.

1984-03-13

A temperature-actuated apparatus is disclosed for releasably supporting a safety rod in a nuclear reactor, comprising a safety rod upper adapter having a retention means, a drive shaft which houses the upper adapter, and a bimetallic means supported within the drive shaft and having at least one ledge which engages a retention means of the safety rod upper adapter. A pre-determined increase in temperature causes the bimetallic means to deform so that the ledge disengages from the retention means, whereby the bimetallic means releases the safety rod into the core of the reactor.

Shielding device for control rod in nuclear reactor

International Nuclear Information System (INIS)

Sakamaki, Kazuo; Tomatsu, Tsutomu.

1995-01-01

The device of the present invention shields radiation emitted from control rods to greatly reduce an operator's radiation exposure even if reactor water level is lowered and the upper portion of the control rod is exposed upon inspection of a BWR type reactor. Namely, a shield assembly has a structure comprising a set of four columnar shields in a two-row and two-column arrangement, which can be inserted into a control rod guide tube. Upon conducting inspection, the control rod is lowered into the control rod guide tube, and in this state, the columnar shields of the shield assembly are inserted to the control rod in the control rod guide tube. With such procedures, the upper portion of the control rod protruded from the control rod guide tube is covered with the shield assembly. As a result, radiation leaked from the control rod is shielded. Accordingly, irradiation in the reactor due to leaked radiation can be prevented thereby enabling to reduce an operator's radiation exposure. (I.S.)

Burnable poison rod

International Nuclear Information System (INIS)

Natsume, Tomohiro.

1988-01-01

Purpose: To increase the reactor core lifetime by decreasing the effect of neutron absorption of burnable poison rods by using material with less neutron absorbing effect. Constitution: Stainless steels used so far as the coating material for burnable poison rods have relatively great absorption in the thermal neutral region and are not preferred in view of the neutron economy. Burnable poison rods having fuel can made of zirconium alloy shows absorption the thermal neutron region lower by one digit than that of stainless steels but they shows absorption in the resonance region and the cost is higher. In view of the above, the fuel can of the burnable poison material is made of aluminum or aluminu alloy. This can reduce the neutron absorbing effect by stainless steel fuel can and effectively utilize neutrons that have been wastefully absorbed and consumed in stainless steels. (Takahashi, M.)

Axial-flow-induced vibration for a rod supported by translational springs at both ends