Evaluation of the oxide and silicide fuels reactivity in the RSG-GAS core

International Nuclear Information System (INIS)

S, Tukiran; M S, Tagor; S, Lily; Pinem, S.

2000-01-01

Fuel exchange of The RSG-GAS reactor core from uranium oxide to uranium silicide in the same loading, density, and enrichment, that is, 250 gr, 2.98 gr/cm 3 , and 19.75 % respectively, will be performed in-step wise. In every cycle of exchange with 5/l mode, it is needed to evaluate the parameter of reactor core operation. One of the important operation parameters is fuel reactivity that gives effect to the core reactivity. The experiment was performed at core no. 36, BOC, low power which exist 2 silicide fuels. The evaluation was done based on the RSG-GAS control rod calibration consisting of 40 fuels and 8 control rod.s. From 40 fuels in the core, there are 2 silicide fuels, RI-225/A-9 and RI-224/C-3. For inserting 2 silicide fuels, the reactivity effect to the core must be know. To know this effect , it was performed fuels reactivity experiment, which based on control rod calibration. But in this case the RSG-GAS has no other fresh oxide fuel so that configuration of the RSG-GAS core was rearranged by taking out the both silicide fuels and this configuration is used as reference core. Then silicide fuel RI-224 was inserted to position F-3 replacing the fresh oxide fuel RI-260 so the different reactivity of the fuels is obtained. The experiment result showed that the fuel reactivity change is in amount of 12.85 cent (0.098 % ) The experiment result was compared to the calculation result, using IAFUEL code which amount to 13.49 cent (0.103 %) The result showed that the reactivity change of oxide to silicide fuel is small so that the fuel exchange from uranium oxide to uranium silicide in the first step can be done without any significant change of the operation parameter

Fast Flux Test Facility (FFTF) feedback reactivity components

International Nuclear Information System (INIS)

Nguyen, D.H.

1988-04-01

The static tests conducted during Cycle 8A (1986) of the FFTF have allowed, for the first time, the experimental determination of each of the feedback reactivities caused by the following mechanisms: fuel axial expansion, control rod repositioning, core radial expansion, and subassembly bowing. A semiempirical equation was obtained to describe each of these feedback components that depended only on the relevant reactor temperature (bowing was presented in a tabular form). The Doppler and sodium density reactivities were calculated using existing mechanistic methods. Although they could also be fitted with closed-form equations depending only on temperatures, these equations are not needed in transient analyses using whole core safety computer codes, which use mechanistic methods. The static feedback reactivity model was extended to obtain a dynamic model via the concept of ''time constants.'' Besides being used for transient analyses in the FFTF, these feedback equations constitute a database for the validation and/or calibration of mechanistic feedback reactivity models. 2 refs., 6 tabs

Magnetic nuclear core restraint and control

International Nuclear Information System (INIS)

Cooper, M.H.

1979-01-01

A lateral restraint and control system for a nuclear reactor core adaptable to provide an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit are composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased a given amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction

Magnetic nuclear core restraint and control

International Nuclear Information System (INIS)

Cooper, M.H.

1979-01-01

A lateral restraint and control systemm for a nuclear reactor core provides an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit is composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased by an amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction

Study on the reactivity behavior partially loaded reactor cores using SIMULATE-3

International Nuclear Information System (INIS)

Holzer, Robert; Zeitz, Andreas; Grimminger, Werner; Lubczyk, Tobias

2009-01-01

The reactor core design for the NPP Gundremmingen unit B and C is performed since several years using the validated 3D reactor core calculation program SIMULATE-3. The authors describe a special application of the program to study the reactivity for different partial core loadings. Based on the comparison with results of the program CASMO-4 the program SIMULATE-3 was validated for the calculation of partially loaded reactor cores. For the planned reactor operation in NPP Gundremmingen using new MOX fuel elements the reactivity behavior was studied with respect to the KTA-Code requirements.

Magnetic nuclear core restraint and control

International Nuclear Information System (INIS)

Cooper, M.H.

1978-01-01

Disclosed is a lateral restraint and control system for a nuclear reactor core adaptable to provide an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit are composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased a given amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction

Higher order polynomial expansion nodal method for hexagonal core neutronics analysis

International Nuclear Information System (INIS)

Jin, Young Cho; Chang, Hyo Kim

1998-01-01

A higher-order polynomial expansion nodal(PEN) method is newly formulated as a means to improve the accuracy of the conventional PEN method solutions to multi-group diffusion equations in hexagonal core geometry. The new method is applied to solving various hexagonal core neutronics benchmark problems. The computational accuracy of the higher order PEN method is then compared with that of the conventional PEN method, the analytic function expansion nodal (AFEN) method, and the ANC-H method. It is demonstrated that the higher order PEN method improves the accuracy of the conventional PEN method and that it compares very well with the other nodal methods like the AFEN and ANC-H methods in accuracy

Moderator temperature effects on reactivity of HEU core of MNSR

International Nuclear Information System (INIS)

Ahmad, Siraj-ul-Islam; Sahibzada, Tasveer Muhammad

2012-01-01

Highlights: ► The MNSR core was analyzed to see the cross section effects on moderator temperature coefficient of reactivity. ► WIMS-D code was used for cell calculations. ► The 3D diffusion theory code PRIDE was first validated using IAEA benchmark problem and then used for analysis of MNSR. ► The differences among results for various libraries were discussed. -- Abstract: In this article we report on analyses that were performed to investigate the influence of cross section differences among libraries released by various centers on reactivity of Miniature Neutron Source Reactors. The 3D model of the core was developed with WIMS-D and PRIDE codes and six cross section libraries were used including JENDL-3.2, JEF-2.2, JEFF-3.3, ENDF/B-VI and ENDF/B-VII, and IAEA library. It was observed that all the libraries predict the reactivity within 10%, with IAEA library giving minimum reactivity worth, and JEF-2.2 data library resulted in highest worth.

Core concept of fast power reactor with zero sodium void reactivity

International Nuclear Information System (INIS)

Matveev, V.I.; Chebeskov, A.N.; Krivitsky, I.Y.

1991-01-01

The paper presents a core concept of BN-800 - type fast power reactor with zero sodium void reactivity (SVR). Consideration is given to the layout-and some design features of such a core. Some considerations on the determination of the required SVR value as one of the fast reactor safety criteria in accidents with coolant boiling are presented. Some methodical considerations an the development of calculation models that give a correct description of the new core features are stated. The results of the integral SVR calculation studies are included. reactivity excursions under different scenarios of sodium boiling are estimated, some corrections into the calculated SVR value are discussed. (author)

Reactivity changes in hybrid thermal-fast reactor systems during fast core flooding

International Nuclear Information System (INIS)

Pesic, M.

1994-09-01

A new space-dependent kinetic model in adiabatic approximation with local feedback reactivity parameters for reactivity determination in the coupled systems is proposed in this thesis. It is applied in the accident calculation of the 'HERBE' fast-thermal reactor system and compared to usual point kinetics model with core-averaged parameters. Advantages of the new model - more realistic picture of the reactor kinetics and dynamics during local large reactivity perturbation, under the same heat transfer conditions, are underlined. Calculated reactivity parameters of the new model are verified in the experiments performed at the 'HERBE' coupled core. The model has shown that the 'HERBE' safety system can shutdown reactor safely and fast even in the case of highly set power trip and even under conditions of big partial failure of the reactor safety system (author)

Fuel density effect on parameter of reactivity coefficient of the Innovative Research Reactor core

International Nuclear Information System (INIS)

Rokhmadi; Tukiran S

2013-01-01

The multipurpose of research reactor utilization make many countries build the new research reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research reactor in Indonesia right now is already 25 year old. Therefor, it is needed to design a new research reactor as a alternative called it innovative research reactor (IRR) and then as an exchanger for old research reactor. The aim of this research is to complete RRI core design data as a requirement for design license. Calculation done is to get the RRI core reactivity coefficients with 5 x 5 core configuration and 20 MW of power, has more than 40 days cycle of length. The RRI core reactivity coefficient calculation is done for new U-"9Mo-Al fuel with variation of densities. The calculation is done by using WIMSD-5B and BATAN-FUEL computer codes. The result of calculation for conceptual design showed that the equilibrium RRI core with 5 x 5 configuration, 450 g, 550 g and 700 g of fuel loadings have negative reactivity coefficients of fuel temperature, moderator temperature, void fraction and density of moderator but the values of the reactivities are very variation. This results has met the safety criteria for RRI core conceptual design. (author)

Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

Science.gov (United States)

Bragg-Sitton, Shannon M.

2005-01-01

The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .tests, the point kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

Three-dimensional core analysis on a super fast reactor with negative local void reactivity

International Nuclear Information System (INIS)

Cao Liangzhi; Oka, Yoshiaki; Ishiwatari, Yuki; Ikejiri, Satoshi

2009-01-01

Keeping negative void reactivity throughout the cycle life is one of the most important requirements for the design of a supercritical water-cooled fast reactor (super fast reactor). Previous conceptual design has negative overall void reactivity. But the local void reactivity, which is defined as the reactivity change when the coolant of one fuel assembly disappears, also needs to be kept negative throughout the cycle life because the super fast reactor is designed with closed fuel assemblies. The mechanism of the local void reactivity is theoretically analyzed from the neutrons balance point of view. Three-dimensional neutronics/thermal-hydraulic coupling calculation is employed to analyze the characteristics of the super fast reactor including the local void reactivity. Some configurations of the core are optimized to decrease the local void reactivity. A reference core is successfully designed with keeping both overall and local void reactivity negative. The maximum local void reactivity is less than -30 pcm

Core expansion in young star clusters in the Large Magellanic Cloud

International Nuclear Information System (INIS)

Elson, R.A.W.; Freeman, K.C.; Lauer, T.R.

1989-01-01

The core radii of 18 rich star clusters in the LMC with ages from 10 Myr to 1 Gyr. Data for an additional 17 clusters with ages from 1 Myr to 10 Gyr are available in the literature. The combined sample shows that the core radii increase from about 0 to about 5 pc between about 1 Myr and 1 Gyr, and then begin to decrease again. The expansion of the cores is probably driven by mass loss from evolving stars. Models of cluster evolution show that the rate of increase in core radius is sensitive to the slope of the initial mass function. The observed core radius-age relation for the LMC clusters favors an intial mass function with slope slightly flatter than the Salpeter value. 20 refs

Alteration of alkali reactive aggregates autoclaved in different alkali solutions and application to alkali-aggregate reaction in concrete (II) expansion and microstructure of concrete microbar

International Nuclear Information System (INIS)

Lu Duyou; Mei Laibao; Xu Zhongzi; Tang Mingshu; Mo Xiangyin; Fournier, Benoit

2006-01-01

The effect of the type of alkalis on the expansion behavior of concrete microbars containing typical aggregate with alkali-silica reactivity and alkali-carbonate reactivity was studied. The results verified that: (1) at the same molar concentration, sodium has the strongest contribution to expansion due to both ASR and ACR, followed by potassium and lithium; (2) sufficient LiOH can completely suppress expansion due to ASR whereas it can induce expansion due to ACR. It is possible to use the duplex effect of LiOH on ASR and ACR to clarify the ACR contribution when ASR and ACR may coexist. It has been shown that a small amount of dolomite in the fine-grained siliceous Spratt limestone, which has always been used as a reference aggregate for high alkali-silica reactivity, might dedolomitize in alkaline environment and contribute to the expansion. That is to say, Spratt limestone may exhibit both alkali-silica and alkali-carbonate reactivity, although alkali-silica reactivity is predominant. Microstructural study suggested that the mechanism in which lithium controls ASR expansion is mainly due to the favorable formation of lithium-containing less-expansive product around aggregate particles and the protection of the reactive aggregate from further attack by alkalis by the lithium-containing product layer

Inadvertent pump start with gas expansion modules

International Nuclear Information System (INIS)

Campbell, L.R.; Harris, R.A.; Heard, F.J.; Dautel, W.A.

1992-01-01

Previous testing demonstrated the effectiveness of gas expansion modules (GEMs) in mitigating the consequences of a loss-of-flow-without-scram transient in Fast Flux Test Facility (FFTF)-sized sodium cooled cores. As a result, GEMs have been included in the advance liquid-metal reactor (PRISM) design project sponsored by the US Department of Energy. The PRISM design is under review at the US Nuclear Regulatory Commission for licensability. In the unlikely event that the reactor does not scram during a loss of low, the GEMs quickly insert sufficient negative reactivity to limit fuel and cladding temperatures to acceptable values. This is the positive benefit of the GEMs; however, the reverse situation must be considered. A primary pump could be inadvertently started from near-critical conditions resulting in a positive reactivity insertion and a power transient. One mitigating aspect of this event is that as the reactivity associated with the GEMs is inserted, the increasing flow increases core cooling. A test was conducted in the FFTF to demonstrate that the GEM and feedback reactivity are well predicted following pump start, and the reactivity transient is benign

Analysis Of Temperature Effects On Reactivity Of The Rsg-Gas Core Using Silicide Fuels

International Nuclear Information System (INIS)

Surbakti, Tukiran; Pinem, Surian

2001-01-01

RSG-GAS has been operating using new silicide fuels so that it is necessary to estimate and to measure the effect of temperature on reactivity of the core. The parameters to be determined due to temperature effect are reactivity coefficient of moderator temperature, temperature coefficient of fuel element and power reactivity coefficient. By doing a couple compensation method, determination of reactivity coefficient as well as the reactivity coefficient of moderator temperature can be obtained. Furthermore, coefficient of the reactivity was successfully estimated using the combination of WIMS-D4 and Batan-2DIFF. The cell calculation was done by using WIMS-D4 code to get macroscopic cross section and Batan-2DIFF code is used for core calculation. The calculation and experimental results of reactivity coefficient do not show any deviation from RSG-GAS safety margin. The results are -2,84 sen/ o C, -1,29 sen/MW and -0,64 sen/ o C for reactivity coefficients of temperature, power, fuel element and moderator temperature, respectively. All of 3 parameters are absolutely met with safety criteria

The influence of spatial effects on the measurement results of reactivity in 'fast disturbances' of core parameters

International Nuclear Information System (INIS)

Tsyganov, S.V.; Shishkov, L.K.

2001-01-01

The analysis of methods for the determination of reactivity revealed an essential influence of spatial effect on the measurement precision. Using of reverse point kinetic equation for reactivity meter is assumed that the average neutron flux weigh with the importance function is known at every moment of the transient. In fact, reactivity meter represent behaviour of the neutron flux only of the part of the core, so measured value of reactivity can differ from really reactivity. Three-dimensional dynamic model of the core allow to evaluate such difference. It is supposed to evaluate correction factor for the neutron flux measured at the place where ion chamber situated with the three-dimensional model NOSTRA of the WWER core. On the basis of such algorithm we propose to build module allowing the influence of spatial effects on the results of the reactivity meter to be eliminated at real time regime. This code will be incorporated into the core monitoring system 'BLOK' (SCORPIO type) which is being developed for the Kola and Rostov NPP. The report illustrates utilization of such algorithm (Authors)

Various reactivity effects value for assuring fast reactor core inherent safety

International Nuclear Information System (INIS)

Belov, S.B.; Vasilyev, B.A.

1991-01-01

The paper presents the results of temperature and power reactivity feedback components calculations for fast reactors with different core volume when using oxide, carbide, nitride and metal fuel. Reactor parameters change in loss of flow without scram and transient over power without scram accidents was evaluated. The importance of various reactivity feedback components in restricting the consequences of these accidents has been analyzed. (author)

Internal core tightener

International Nuclear Information System (INIS)

Brynsvold, G.V.; Snyder, H.J. Jr.

1976-01-01

An internal core tightener is disclosed which is a linear actuated (vertical actuation motion) expanding device utilizing a minimum of moving parts to perform the lateral tightening function. The key features are: (1) large contact areas to transmit loads during reactor operation; (2) actuation cam surfaces loaded only during clamping and unclamping operation; (3) separation of the parts and internal operation involved in the holding function from those involved in the actuation function; and (4) preloaded pads with compliant travel at each face of the hexagonal assembly at the two clamping planes to accommodate thermal expansion and irradiation induced swelling. The latter feature enables use of a ''fixed'' outer core boundary, and thus eliminates the uncertainty in gross core dimensions, and potential for rapid core reactivity changes as a result of core dimensional change. 5 claims, 12 drawing figures

The reactivity meter and core reactivity

International Nuclear Information System (INIS)

Siltanen, P.

1999-01-01

This paper discussed in depth the point kinetic equations and the characteristics of the point kinetic reactivity meter, particularly for large negative reactivities. From a given input signal representing the neutron flux seen by a detector, the meter computes a value of reactivity in dollars (ρ/β), based on inverse point kinetics. The prompt jump point of view is emphasised. (Author)

Combination of AC Transmission Expansion Planning and Reactive Power Planning in the restructured power system

International Nuclear Information System (INIS)

Hooshmand, Rahmat-Allah; Hemmati, Reza; Parastegari, Moein

2012-01-01

Highlights: ► To overcome the disadvantages of DC model in Transmission Expansion Planning, AC model should be used. ► The Transmission Expansion Planning associated with Reactive Power Planning results in fewer new transmission lines. ► Electricity market concepts should be considered in Transmission Expansion Planning problem. ► Reliability aspects should be considered in Transmission Expansion Planning problem. ► Particle Swarm Optimization is a suitable optimization method to solve Transmission Expansion Planning problem. - Abstract: Transmission Expansion Planning (TEP) is an important issue in power system studies. It involves decisions on location and number of new transmission lines. Before deregulation of the power system, the goal of TEP problem was investment cost minimization. But in the restructured power system, nodal prices, congestion management, congestion surplus and so on, have been considered too. In this paper, an AC model of TEP problem (AC-TEP) associated with Reactive Power Planning (RPP) is presented. The goals of the proposed planning problem are to minimize investment cost and maximize social benefit at the same time. In the proposed planning problem, in order to improve the reliability of the system the Expected Energy Not Supplied (EENS) index of the system is limited by a constraint. For this purpose, Monte Carlo simulation method is used to determine the EENS. Particle Swarm Optimization (PSO) method is used to solve the proposed planning problem which is a nonlinear mixed integer optimization problem. Simulation results on Garver and RTS systems verify the effectiveness of the proposed planning problem for reduction of the total investment cost, EENS index and also increasing social welfare of the system.

Analysis of excess reactivity of JOYO MK-III performance test core

International Nuclear Information System (INIS)

Maeda, Shigetaka; Yokoyama, Kenji

2003-10-01

JOYO is currently being upgraded to the high performance irradiation bed JOYO MK-III core'. The MK-III core is divided into two fuel regions with different plutonium contents. To obtain a higher neutron flux, the active core height was reduced from 55 cm to 50 cm. The reflector subassemblies were replaced by shielding subassemblies in the outer two rows. Twenty of the MK-III outer core fuel subassemblies in the performance test core were partially burned in the transition core. Four irradiation test rigs, which do not contain any fuel material, were loaded in the center of the performance test core. In order to evaluate the excess reactivity of MK-III performance test core accurately, we evaluated it by applying not only the JOYO MK-II core management code system MAGI, but also the MK-III core management code system HESTIA, the JUPITER standard analysis method and the Monte Carlo method with JFS-3-J3.2R content set. The excess reactivity evaluations obtained by the JUPITER standard analysis method were corrected to results based on transport theory with zero mesh-size in space and angle. A bias factor based on the MK-II 35th core, which sensitivity was similar to MK-III performance test core's, was also applied, except in the case where an adjusted nuclear cross-section library was used. Exact three-dimensional, pin-by-pin geometry and continuous-energy cross sections were used in the Monte Carlo calculation. The estimated error components associated with cross-sections, methods correction factors and the bias factor were combined based on Takeda's theory. Those independently calculated values agree well and range from 2.8 to 3.4%Δk/kk'. The calculation result of the MK-III core management code system HESTLA was 3.13% Δk/kk'. The estimated errors for bias method range from 0.1 to 0.2%Δk/kk'. The error in the case using adjusted cross-section was 0.3%Δk/kk'. (author)

Reactivity considerations for the on-line refuelling of a pebble bed modular reactor—Illustrating safety for the most reactive core fuel load

International Nuclear Information System (INIS)

Reitsma, Frederik

2012-01-01

In the multi-pass fuel management scheme employed for the pebble bed modular reactor the fuel pebbles are re-circulated until they reach the target burn-up. The rate at which fresh fuel is loaded and burned fuel is discharged is a result of the core neutronics cycle analysis but in practice (on the plant) this has to be controlled and managed by the fuel handling and storage system and use of the burnup measurement system. The excess reactivity is the additional reactivity available in the core during operating conditions that is the result of loading a fuel mixture in the core that is more reactive (less burned) than what is required to keep the reactor critical at full power operational conditions. The excess reactivity is balanced by the insertion of the control rods to keep the reactor critical. The excess reactivity allows flexibility in operations, for example to overcome the xenon build up when power is decreased as part of load follow. In order to limit reactivity excursions and to ensure safe shutdown the excess reactivity and thus the insertion depth of the control rods at normal operating conditions has to be managed. One way to do this is by operational procedures. The reactivity effect of long-term operation with the control rods inserted deeper than the design point is investigated and a control rod insertion limit is proposed that will not limit normal operations. The effects of other phenomena that can increase the power defect, such as higher-than-expected fuel temperatures, are also introduced. All of these cases are then evaluated by ensuring cold shutdown is still achievable and where appropriate by reactivity insertion accident analysis. These aspects are investigated on the PBMR 400 MW design.

Analysis of reactivity accidents of the RSG-GAS core with silicide fuel

International Nuclear Information System (INIS)

Tukiran

2002-01-01

The fuels of RSG-GAS reactor is changed from uranium oxide to uranium silicide. For time being, the fuel of RSG-GAS core are mixed up between oxide and silicide fuels with 250 gr of loading and 2.96 g U/cm 3 of density, respectively. While, silicide fuel with 300 gr of loading is still under research. The advantages of silicide fuels are can be used in high density, so that, it can be stayed longer in the core at higher burn-up, therefore, the length of cycle is longer. The silicide fuel in RSG-GAS core is used in step-wise by using mixed up core. Firstly, it is used silicide fuel with 250 gr of loading and then, silicide fuel with 300 gr of loading (3.55 g U/cm 3 of density). In every step-wise of fuel loading must be analysed its safety margin. In this occasion, it is analysed the reactivity accident of RSG-GAS core with 300 gr of silicide fuel loading. The calculation was done by using POKDYN code which available at P2TRR. The calculation was done by reactivity insertion at start up and power rangers. From all cases which were have been done, the results of analysis showed that there is no anomaly and safety margin break at RSG-GAS core with 300 gr silicide fuel loading

Effects of conversion ratio change on the core performances in medium to large TRU burning reactors

International Nuclear Information System (INIS)

Song, Hoon; Kim, Sang-Ji; Yoo, Jae-Woon; Kim, Yeong-Il

2009-01-01

Conceptual fast reactor core designs with sodium coolant are developed at 1,500, 3,000 and 4,500 MWt which are configured to transmute recycled transuranics (TRU) elements with external feeds consisting of LWR spent fuel. Even at each pre-determined power level, the performance parameters, reactivity coefficients and their implications on the safety analysis can be different when the target TRU conversion ratio changes. In order to address this aspect of design, a study on TRU conversion ratio change was performed. The results indicate that it is feasible to design a TRU burner core to accommodate a wide range of conversion ratios by employing different fuel cladding thicknesses. The TRU consumption rate is found to be proportional to the core power without any significant deterioration in the core performance at higher power levels. A low conversion ratio core has an increased TRU consumption rate and much faster burnup reactivity loss, which calls for appropriate means for reactivity compensation. As for the reactivity coefficients related with the conversion ratio change, the core with a low conversion ratio has a less negative Doppler coefficient, a more negative axial expansion coefficient, a more negative control rod worth per rod, a more negative radial expansion coefficient, a less positive sodium density coefficient and a less positive sodium void worth. A slight decrease in the delayed neutron fraction is also noted, reflecting the fertile U-238 fraction reduction. (author)

Measurement and analysis of reactivity worth of 237Np sample in cores of TCA and FCA

International Nuclear Information System (INIS)

Sakurai, Takeshi; Mori, Takamasa; Okajima, Shigeaki; Tani, Kazuhiro; Suzaki, Takenori; Saito, Masaki

2009-01-01

The reactivity worth of 22.87 grams of 237 Np oxide sample was measured and analyzed in seven uranium cores in the Tank-Type Critical Assembly (TCA) and two uranium cores in the Fast Critical Assembly (FCA) at the Japan Atomic Energy Agency. The TCA cores provided a systematic variation in the neutron spectrum between the thermal and resonance energy regions. The FCA cores, XXI and XXV, provided a hard neutron spectrum of the fast reactor and a soft one of the resonance energy region, respectively. Analyses were carried out using the JENDL-3.3 nuclear data library with a Monte Carlo method for the TCA cores and a deterministic method for the FCA cores. The ratios of calculated to experimental (C/E) reactivity worth were between 0.97 and 0.91, and showed no apparent dependence on the neutron spectrum. (author)

Development and Investigation of Reactivity Measurement Methods in Subcritical Cores

Energy Technology Data Exchange (ETDEWEB)

Wright, Johanna

2005-05-01

Subcriticality measurements during core loading and in future accelerator driven systems have a clear safety relevance. In this thesis two subcriticality methods are treated: the Feynman-alpha and the source modulation method. The Feynman-alpha method is a technique to determine the reactivity from the relative variance of the detector counts during a measurement period. The period length is varied to get the full time dependence of the variance-to-mean. The corresponding theoretical formula was known only with stationary sources. In this thesis, due to its relevance for novel reactivity measurement methods, the Feynman-alpha formulae for pulsed sources for both the stochastic and the deterministic cases are treated. Formulae neglecting as well as including the delayed neutrons are derived. The formulae neglecting delayed neutrons are experimentally verified with quite good agreement. The second reactivity measurement technique investigated in this thesis is the so-called source modulation technique. The theory of the method was elaborated on the assumption of point kinetics, but in practice the method will be applied by using the signal from a single local neutron detector. Applicability of the method therefore assumes point kinetic behaviour of the core. Hence, first the conditions of the point kinetic behaviour of subcritical cores was investigated. After that the performance of the source modulation technique in the general case as well as and in the limit of exact point kinetic behaviour was examined. We obtained the unexpected result that the method has a finite, non-negligible error even in the limit of point kinetic behaviour, and a substantial error in the operation range of future accelerator driven subcritical reactors (ADS). In practice therefore the method needs to be calibrated by some other method for on-line applications.

Development and Investigation of Reactivity Measurement Methods in Subcritical Cores

International Nuclear Information System (INIS)

Wright, Johanna

2005-05-01

Subcriticality measurements during core loading and in future accelerator driven systems have a clear safety relevance. In this thesis two subcriticality methods are treated: the Feynman-alpha and the source modulation method. The Feynman-alpha method is a technique to determine the reactivity from the relative variance of the detector counts during a measurement period. The period length is varied to get the full time dependence of the variance-to-mean. The corresponding theoretical formula was known only with stationary sources. In this thesis, due to its relevance for novel reactivity measurement methods, the Feynman-alpha formulae for pulsed sources for both the stochastic and the deterministic cases are treated. Formulae neglecting as well as including the delayed neutrons are derived. The formulae neglecting delayed neutrons are experimentally verified with quite good agreement. The second reactivity measurement technique investigated in this thesis is the so-called source modulation technique. The theory of the method was elaborated on the assumption of point kinetics, but in practice the method will be applied by using the signal from a single local neutron detector. Applicability of the method therefore assumes point kinetic behaviour of the core. Hence, first the conditions of the point kinetic behaviour of subcritical cores was investigated. After that the performance of the source modulation technique in the general case as well as and in the limit of exact point kinetic behaviour was examined. We obtained the unexpected result that the method has a finite, non-negligible error even in the limit of point kinetic behaviour, and a substantial error in the operation range of future accelerator driven subcritical reactors (ADS). In practice therefore the method needs to be calibrated by some other method for on-line applications

TRACE analysis of Phenix core response to an increase of the core inlet sodium temperature

Energy Technology Data Exchange (ETDEWEB)

Chenu, A., E-mail: aurelia.chenu@psi.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Ecole Polytechnique Federale (Switzerland); Mikityuk, K., E-mail: konstantin.mikityuk@psi.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Adams, R., E-mail: robert.adams@psi.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Eidgenossische Technische Hochschule, Zurich (Switzerland); Chawla, R., E-mail: rakesh.chawla@epfl.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Ecole Polytechnique Federale (Switzerland)

2011-07-01

This work presents the analysis, using the TRACE code, of the Phenix core response to an inlet sodium temperature increase. The considered experiment was performed in the frame of the Phenix End-Of-Life (EOL) test program of the CEA, prior to the final shutdown of the reactor. It corresponds to a transient following a 40°C increase of the core inlet temperature, which leads to a power decrease of 60%. This work focuses on the first phase of the transient, prior to the reactor scram and pump trip. First, the thermal-hydraulic TRACE model of the core developed for the present analysis is described. The kinetic parameters and feedback coefficients for the point kinetic model were first derived from a 3D static neutronic ERANOS model developed in a former study. The calculated kinetic parameters were then optimized, before use, on the basis of the experimental reactivity in order to minimize the error on the power calculation. The different reactivity feedbacks taken into account include various expansion mechanisms that have been specifically implemented in TRACE for analysis of fast-neutron spectrum systems. The point kinetic model has been used to study the sensitivity of the core response to the different feedback effects. The comparison of the calculated results with the experimental data reveals the need to accurately calculate the reactivity feedback coefficients. This is because the reactor response is very sensitive to small reactivity changes. This study has enabled us to study the sensitivity of the power change to the different reactivity feedbacks and define the most important parameters. As such, it furthers the validation of the FAST code system, which is being used to gain a more in-depth understanding of SFR core behavior during accidental transients. (author)

Development of a perturbation code, PERT-K, for hexagonal core geometry

Energy Technology Data Exchange (ETDEWEB)

Kim, Taek Kyum; Kim, Sang Ji; Song, Hoon; Kim, Young Il; Kim, Young Jin [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-01-01

A perturbation code for hexagonal core geometry has been developed based on Nodal Expansion Method. By using relevant output files of DIF3D code, it can calculate the reactivity changes caused by perturbation in composition or/and neutron cross section libraries. The accuracy of PERT-K code has been validated by calculating the reactivity changes due to fuel composition change, the sodium void coefficients, and the sample reactivity worths of BFS-73-1 critical experiments. In the case of 10% reduction in all fuel isotopics at a assembly located in the outer core, PERT-K computation agrees with the direct computation by DIF3D within 60 pcm. The sample reactivity worths of BFS-73-1 critical experiments are predicted with PERT-K code within the experimental error bounds. For 100% sodium void occurrence at the inner core, the maximum difference of reactivity changes between PERT-K and direct DIF3D computations is less than 40 pcm. On the other hand, the same sodium void condition at the outer core leads to a difference of reactivity change greater than 400 pcm. However, as sodium voiding becomes near zero value, the difference becomes less and rapidly falls within the acceptable bound, i.e. 40 pcm. (author). 11 refs., 9 figs., 6 tabs.

Impact of correlations between core configurations for the evaluation of nuclear data uncertainty propagation for reactivity

International Nuclear Information System (INIS)

Frosio, T.; Bonaccorsi, T.; Blaise, P.

2017-01-01

The precise estimation of Pearson correlations, also called 'representativity' coefficients, between core configurations is a fundamental quantity for properly assessing the nuclear data (ND) uncertainties propagation on integral parameters such as k-eff, power distributions, or reactivity coefficients. In this paper, a traditional adjoint method is used to propagate ND uncertainty on reactivity and reactivity coefficients and estimate correlations between different states of the core. We show that neglecting those correlations induces a loss of information in the final uncertainty. We also show that using approximate values of Pearson does not lead to an important error of the model. This calculation is made for reactivity at the beginning of life and can be extended to other parameters during depletion calculations. (authors)

A technique for computing bowing reactivity feedback in LMFBR's

International Nuclear Information System (INIS)

Finck, P.J.

1987-01-01

During normal or accidental transients occurring in a LMFBR core, the assemblies and their support structure are subjected to important thermal gradients which induce differential thermal expansions of the walls of the hexcans and differential displacement of the assembly support structure. These displacements, combined with the creep and swelling of structural materials, remain quite small, but the resulting reactivity changes constitute a significant component of the reactivity feedback coefficients used in safety analyses. It would be prohibitive to compute the reactivity changes due to all transients. Thus, the usual practice is to generate reactivity gradient tables. The purpose of the work presented here is twofold: develop and validate an efficient and accurate scheme for computing these reactivity tables; and to qualify this scheme

Track 5: safety in engineering, construction, operations, and maintenance. Reactor physics design, validation, and operating experience. 5. A Negative Reactivity Feedback Device for Actinide Burner Cores

International Nuclear Information System (INIS)

Driscoll, M.J.; Hejzlar, P.

2001-01-01

Lead-bismuth eutectic (LBE) cooled reactors are of considerable interest because they may be useful for destruction of actinides in a cost-effective manner, particularly cores fueled predominantly with minor actinides, which gain reactivity with burnup. However, they also pose several design challenges: 1. a small (and perhaps even slightly positive) Doppler feedback; 2. small effective delayed neutron yield; 3. a small negative feedback from axial fuel expansion; 4. positive coolant void and temperature coefficients for conventional designs. This has motivated a search for palliative measures, leading to conceptualization of the reactivity feedback device (RFD). The RFD consists of an in-core flask containing helium gas, tungsten wool, and a small reservoir of LBE that communicates with vertical tubes housing neutron absorber floats. The upper part of these guide tubes contains helium gas that is vented into a separate, cooler ex-core helium gas plenum. The principle of operation is as follows: 1. The tungsten wool, hence the helium gas in the in-core plenum, is heated by gammas and loses heat to the walls by convection and conduction (radiation is feeble for monatomic gases and, in any event, intercepted by the tungsten wool). An energy balance determines the gas temperature, hence, pressure, which is 10 atm here. The energy loss rate can be adjusted by using xenon or a gas mixture in place of helium. The tungsten wool mass, which is 1 vol% wool here, can also be increased to increase gamma heating and further retard convection; alternatively, a Dewar flask could be used in place of the additional wool. 2. An increase in core power causes a virtually instantaneous increase in gamma flux, hence, gas heatup: The thermal time constant of the tungsten filaments and their surrounding gas film is ∼40 μs. 3. The increased gas temperature is associated with an increased gas pressure, which forces more liquid metal into the float guide tubes: LBE will rise ∼100 cm

Sodium-cooled fast reactor core designs for transmutation of MHR spent fuel

International Nuclear Information System (INIS)

Hong, S. G.; Kim, Y. H.; Venneri, F.

2010-01-01

In this paper, the core design analyses of sodium cooled fast reactors (SFR) are performed for the effective transmutation of the DB (Deep Burn)-MHR (Modular Helium Reactor). In this concept, the spent fuels of DB-MHR are transmuted in SFRs with a closed fuel cycle after TRUs from LWR are first incinerated in a DB-MHR. We introduced two different type SFR core designs for this purpose, and evaluated their core performance parameters including the safety-related parameters. In particular, the cores are designed to have lower transmutation rate relatively to our previous work so as to make the fuel characteristics more feasible. The first type cores which consist of two enrichment regions are typical homogeneous annular cores and they rate 900 MWt power. On the other hand, the second type cores which consist of a central non-fuel region and a single enrichment fuel region rate relatively higher power of 1500 MWt. For these cores, the moderator rods (YH 1.8 ) are used to achieve less positive sodium void worth and the more negative Doppler coefficient because the loading of DB-MHR spent fuel leads to the degradation of these safety parameters. The analysis results show that these cores have low sodium void worth and negative reactivity coefficients except for the one related with the coolant expansion but the coolant expansion reactivity coefficient is within the typical range of the typical SFR cores. (authors)

Thermal-hydraulics and neutronics studies on the FP7 CP-ESFR oxide and carbide cores

Energy Technology Data Exchange (ETDEWEB)

Ammirabile, L.; Tsige-Tamirat, H. [European Commission, JRC, Inst. for Energy, Petten (Netherlands)

2011-07-01

In the framework of the the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) two core designs that are currently being proposed for the 3600 MWth sodium-cooled reactor concept: one is based on oxide fuel and the other on carbide fuel. Using the European Safety Assessment Platform (ESAP), JRC-IE has conducted static calculation on neutronics (incl. reactivity coefficients) and thermal-hydraulic characteristics for both oxide and carbide reference cores. The quantities evaluated include: keff, coolant heat-up, void, and Doppler reactivity coefficients, axial and radial expansion reactivity coefficients, pin-by-pin calculated power profiles, average and peak channel temperatures. This paper presents the ESAP models applied in the study together with the relevant results for the oxide and carbide core. (author)

Thermal-hydraulics and neutronics studies on the FP7 CP-ESFR oxide and carbide cores

International Nuclear Information System (INIS)

Ammirabile, L.; Tsige-Tamirat, H.

2011-01-01

In the framework of the the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) two core designs that are currently being proposed for the 3600 MWth sodium-cooled reactor concept: one is based on oxide fuel and the other on carbide fuel. Using the European Safety Assessment Platform (ESAP), JRC-IE has conducted static calculation on neutronics (incl. reactivity coefficients) and thermal-hydraulic characteristics for both oxide and carbide reference cores. The quantities evaluated include: keff, coolant heat-up, void, and Doppler reactivity coefficients, axial and radial expansion reactivity coefficients, pin-by-pin calculated power profiles, average and peak channel temperatures. This paper presents the ESAP models applied in the study together with the relevant results for the oxide and carbide core. (author)

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

International Nuclear Information System (INIS)

Mohamed, F.M.

2011-01-01

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

Benchscale Assessment of the Efficacy of a Reactive Core Mat to Isolate PAH-spiked Aquatic Sediments.

Science.gov (United States)

Meric, Dogus; Barbuto, Sara; Sheahan, Thomas C; Shine, James P; Alshawabkeh, Akram N

2014-01-01

This paper describes the results of a benchscale testing program to assess the efficacy of a reactive core mat (RCM) for short term isolation and partial remediation of contaminated, subaqueous sediments. The 1.25 cm thick RCM (with a core reactive material such as organoclay with filtering layers on top and bottom) is placed on the sediment, and approximately 7.5 - 10 cm of overlying soil is placed on the RCM for stability and protection. A set of experiments were conducted to measure the sorption characteristics of the mat core (organoclay) and sediment used in the experiments, and to determine the fate of semi-volatile organic contaminants and non-reactive tracers through the sediment and reactive mat. The experimental study was conducted on naphthalene-spiked Neponset River (Milton, MA) sediment. The results show nonlinear sorption behavior for organoclay, with sorption capacity increasing with increasing naphthalene concentration. Neponset River sediment showed a notably high sorption capacity, likely due to the relatively high organic carbon fraction (14%). The fate and transport experiments demonstrated the short term efficiency of the reactive mat to capture the contamination that is associated with the post-capping period during which the highest consolidation-induced advective flux occurs, driving solid particles, pore fluid and soluble contaminants toward the reactive mat. The goal of the mat placement is to provide a physical filtering and chemically reactive layer to isolate contamination from the overlying water column. An important finding is that because of the high sorption capacity of the Neponset River sediment, the physical filtering capability of the mat is as critical as its chemical reactive capacity.

Influence of external source location in the reactivity calculation

International Nuclear Information System (INIS)

Silva, Adilson Costa da; Silva, Fernando Carvalho da; Martinez, Aquilino Senra

2011-01-01

We used the neutron diffusion equation with external neutron sources, in cartesian geometry and the two groups of energy, to verify the influence of external neutron source locations in the reactivity calculation. For this, a coarse mesh finite difference method was developed for the adjoint flux calculation and simplifies reactivity calculation in PWR type reactor, which uses the output of the nodal expansion method. The results were obtained for different locations on the two-dimensional plane, as well as for different types of fuel elements in the reactor core. (author)

Influence of external source location in the reactivity calculation

Energy Technology Data Exchange (ETDEWEB)

Silva, Adilson Costa da; Silva, Fernando Carvalho da; Martinez, Aquilino Senra, E-mail: asilva@con.ufrj.b, E-mail: fernando@con.ufrj.b, E-mail: Aquilino@lmp.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

2011-07-01

We used the neutron diffusion equation with external neutron sources, in cartesian geometry and the two groups of energy, to verify the influence of external neutron source locations in the reactivity calculation. For this, a coarse mesh finite difference method was developed for the adjoint flux calculation and simplifies reactivity calculation in PWR type reactor, which uses the output of the nodal expansion method. The results were obtained for different locations on the two-dimensional plane, as well as for different types of fuel elements in the reactor core. (author)

EXCURS: a computing programme for analysis of core transient behaviour in a sodium cooled fast reactor

International Nuclear Information System (INIS)

Saito, Shinzo

1977-09-01

In the code EXCURS developed for core transient behaviour calculation of a sodium-cooled fast reactor, a one-channel model is used to represent thermal behaviour of the reactor core. Calculations are made for three different channels; i.e. average, hot and hottest. In the average channel the power density and coolant velocity are equal to the mean values of the whole core. In the hot channel, a maximum power density of the core and a specific coolant velocity are introduced. In the hottest channel, engineering hot channel factors are considered to the hot channel. A one-point neutron kinetics equation with six delayed neutron groups is used to calculate the time-dependent power behaviour. Externally introduced reactivity effect and control rod movement in the case of a scram are taken into account. In the feedback effects evaluated on the basis of the average channel temperatures are considered Doppler effect, fuel axial expansion, cladding expansion, coolant expansion and structure expansion. The decay heat after reactor scram is also considered. Heat balance is taken in each cross section, neglecting the axial heat transfer except for the coolant region. Temperature dependence of the physical properties of materials is considered by second-order polynomials approximation, and also the fuel melting process. Each channel can be divided into a maximum of 20 regions in both radially and axially. The reactor core transient behaviour due to reactivity insertion or loss-of-coolant flow can be studied by EXCURS. The calculated results are plotted optionally by connected code EXPLOT. (auth.)

Analytic function expansion nodal method for nuclear reactor core design

International Nuclear Information System (INIS)

Noh, Hae Man

1995-02-01

In most advanced nodal methods the transverse integration is commonly used to reduce the multi-dimensional diffusion equation into equivalent one- dimensional diffusion equations when derving the nodal coupling equations. But the use of the transverse integration results in some limitations. The first limitation is that the transverse leakage term which appears in the transverse integration procedure must be appropriately approximated. The second limitation is that the one-dimensional flux shapes in each spatial direction resulted from the nodal calculation are not accurate enough to be directly used in reconstructing the pinwise flux distributions. Finally the transverse leakage defined for a non-rectangular node such as a hexagonal node or a triangular node is too complicated to be easily handled and may contain non-physical singular terms of step-function and delta-function types. In this thesis, the Analytic Function Expansion Nodal (AFEN) method and its two variations : the Polynomial Expansion Nodal (PEN) method and the hybrid of the AFEN and PEN methods, have been developed to overcome the limitations of the transverse integration procedure. All of the methods solve the multidimensional diffusion equation without the transverse integration. The AFEN method which we believe is the major contribution of this study to the reactor core analysis expands the homogeneous flux distributions within a node in non-separable analytic basis functions satisfying the neutron diffusion equations at any point of the node and expresses the coefficients of the flux expansion in terms of the nodal unknowns which comprise a node-average flux, node-interface fluxes, and corner-point fluxes. Then, the nodal coupling equations composed of the neutron balance equations, the interface current continuity equations, and the corner-point leakage balance equations are solved iteratively to determine all the nodal unknowns. Since the AFEN method does not use the transverse integration in

Effects of moderation level on core reactivity and. neutron fluxes in natural uranium fueled and heavy water moderated reactors

International Nuclear Information System (INIS)

Khan, M.J.; Aslam; Ahmad, N.; Ahmed, R.; Ahmad, S.I.

2005-01-01

The neutron moderation level in a nuclear reactor has a strong influence on core multiplication, reactivity control, fuel burnup, neutron fluxes etc. In the study presented in this article, the effects of neutron moderation level on core reactivity and neutron fluxes in a typical heavy water moderated nuclear research reactor is explored and the results are discussed. (author)

Spontaneous stabilization of HTGRs without reactor scram and core cooling—Safety demonstration tests using the HTTR: Loss of reactivity control and core cooling

Energy Technology Data Exchange (ETDEWEB)

Takamatsu, Kuniyoshi, E-mail: takamatsu.kuniyoshi@jaea.go.jp; Yan, Xing L.; Nakagawa, Shigeaki; Sakaba, Nariaki; Kunitomi, Kazuhiko

2014-05-01

It is well known that a High-Temperature Gas-cooled Reactor (HTGR) has superior safety characteristics; for example, an HTGR has a self-control system that uses only physical phenomena against various accidents. Moreover, the large heat capacity and low power density of the core result in very slow temperature transients. Therefore, an HTGR serves inherently safety features against loss of core cooling accidents such as the Tokyo Electric Power Co., Inc. (TEPCO)’s Fukushima Daiichi Nuclear Power Station (NPS) disaster. Herein we would like to demonstrate the inherent safety features using the High-Temperature Engineering Test Reactor (HTTR). The HTTR is the first HTGR in Japan with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C; it was built at the Oarai Research and Development Center of Japan Atomic Energy Agency (JAEA). In this study, an all-gas-circulator trip test was analyzed as a loss of forced cooling (LOFC) test with an initial reactor power of 9 MW to demonstrate LOFC accidents. The analytical results indicate that reactor power decreases from 9 MW to 0 MW owing to the negative reactivity feedback effect of the core, even if the reactor shutdown system is not activated. The total reactivity decreases for 2–3 h and then gradually increases in proportion to xenon reactivity; therefore, the HTTR achieves recritical after an elapsed time of 6–7 h, which is different from the elapsed time at reactor power peak occurrence. After the reactor power peak occurs, the total reactivity oscillates several times because of the negative reactivity feedback effect and gradually decreases to zero. Moreover, the new conclusions are as follows: the greater the amount of residual heat removed from the reactor core, the larger the stable reactor power after recriticality owing to the heat balance of the reactor system. The minimum reactor power and the reactor power peak occurrence are affected by the neutron source. The greater the

Comparison of the behaviour of two core designs for ASTRID in case of severe accidents

Energy Technology Data Exchange (ETDEWEB)

Bertrand, F., E-mail: frederic.bertrand@cea.fr [CEA, DEN, DER, F-13108 Saint Paul-lez-Durance (France); Marie, N.; Prulhière, G.; Lecerf, J. [CEA, DEN, DER, F-13108 Saint Paul-lez-Durance (France); Seiler, J.M. [CEA, DEN, DTN, F-38054 Grenoble (France)

2016-02-15

Highlights: • Low void worth CFV and SFRv2 cores are compared for ASTRID pre-conceptual design. • Severe accident behaviour is assessed with a simplified calculation approach and tools. • Mitigation to limit reactivity inserted by core compaction is easier for CFV than for SFRv2 core. • When facing arbitrary reactivity ramps, CFV core would lead to lower energy release than SFRv2 core. • Time scale for core degradation is one order of magnitude larger for CFV than for SFRv2. - Abstract: The present paper is dedicated to the studies carried out during the first stage of the pre-conceptual design of the French demonstrator of fourth generation SFR reactors (ASTRID) in order to compare the behaviour of two envisaged core concepts under severe accident transients. Among the two studied core concepts, whose powers are 1500 MWth, the first one is a classical homogeneous core (called SFRv2) with large pin diameter whose the sodium overall voiding reactivity effect is 5 $. The second concept is an axially heterogeneous core (called CFV) whose global void reactivity effect is negative (−1.2 $ at the end of cycle at the equilibrium). The comparison of the cores relies on two typical accident families: a reactivity insertion (unprotected transient overpower, UTOP) and an overall loss of core cooling (unprotected loss of flow, ULOF). In the first part of the comparison, the primary phase of an UTOP is studied in order to assess typical features of the transient behaviour: power and reactivity evolutions, material heating and melting/vaporization and mechanical energy release due to fuel vapor expansion. The second part of the comparison deals with the calculation of the reactivity potential for degraded states (molten pools) representative of the secondary phase of a mild UTOP and of a strong UTOP (strong or mild qualifies the reactivity ramp inserted). According to the reactivity potential, the amount of fuel to extract from the core and the amount of absorber

HELIOS/DRAGON/NESTLE codes' simulation of void reactivity in a CANDU core

International Nuclear Information System (INIS)

Sarsour, H.N.; Rahnema, F.; Mosher, S.; Turinsky, P.J.; Serghiuta, D.; Marleau, G.; Courau, T.

2002-01-01

This paper presents results of simulation of void reactivity in a CANDU core using the NESTLE core simulator, cross sections from the HELIOS lattice physics code in conjunction with incremental cross sections from the DRAGON lattice physics code. First, a sub-region of a CANDU6 core is modeled using the NESTLE core simulator and predictions are contrasted with predictions by the MCNP Monte Carlo simulation code utilizing a continuous energy model. In addition, whole core modeling results are presented using the NESTLE finite difference method (FDM), NESTLE nodal method (NM) without assembly discontinuity factors (ADF), and NESTLE NM with ADF. The work presented in this paper has been performed as part of a project sponsored by the Canadian Nuclear Safety Commission (CNSC). The purpose of the project was to gather information and assess the accuracy of best estimate methods using calculational methods and codes developed independently from the CANDU industry. (author)

Effects of Radial Reflector Composition on Core Reactivity and Peak Power

International Nuclear Information System (INIS)

Park, Sang Yoon; Lee, Kyung Hoon; Song, Jae Seung

2007-10-01

The effects of radial SA-240 alloy shroud on core reactivity and peak power are evaluated. The existence of radial SA-240 alloy shroud makes reflector water volume decrease, so the thermal absorption cross section of radial reflector is lower than without SA-240 alloy shroud case. Finally, the cycle length is increased from 788 EFPD to 845 EFPD and the peak power is decreased from 1.66 to 1.49. In the case of without SA-240 alloy shroud, a new core loading pattern search has been performed. For the guarantee of the same equivalent cycle length of with SA-240 alloy shroud case, the enrichment of U-235 should be increased from 4.22 w/o to 4.68 w/o. The nuclear key safety parameters of new core loading pattern have been calculated and recorded for the future

Effects of Radial Reflector Composition on Core Reactivity and Peak Power

Energy Technology Data Exchange (ETDEWEB)

Park, Sang Yoon; Lee, Kyung Hoon; Song, Jae Seung

2007-10-15

The effects of radial SA-240 alloy shroud on core reactivity and peak power are evaluated. The existence of radial SA-240 alloy shroud makes reflector water volume decrease, so the thermal absorption cross section of radial reflector is lower than without SA-240 alloy shroud case. Finally, the cycle length is increased from 788 EFPD to 845 EFPD and the peak power is decreased from 1.66 to 1.49. In the case of without SA-240 alloy shroud, a new core loading pattern search has been performed. For the guarantee of the same equivalent cycle length of with SA-240 alloy shroud case, the enrichment of U-235 should be increased from 4.22 w/o to 4.68 w/o. The nuclear key safety parameters of new core loading pattern have been calculated and recorded for the future.

Automatic determination of pressurized water reactor core loading patterns that maximize beginning-of-cycle reactivity within power-peaking and burnup constraints

International Nuclear Information System (INIS)

Hobson, G.H.; Turinsky, P.J.

1986-01-01

Computational capability has been developed to automatically determine a good estimate of the core loading pattern, which minimizes fuel cycle costs for a pressurized water reactor (PWR). Equating fuel cycle cost minimization with core reactivity maximization, the objective is to determine the loading pattern that maximizes core reactivity while satisfying power peaking, discharge burnup, and other constraints. The method utilizes a two-dimensional, coarse-mesh, finite difference scheme to evaluate core reactivity and fluxes for an initial reference loading pattern. First-order perturbation theory is applied to determine the effects of assembly shuffling on reactivity, power distribution, end-of-cycle burnup. Monte Carlo integer programming is then used to determine a near-optimal loading pattern within a range of loading patterns near the reference pattern. The process then repeats with the new loading pattern as the reference loading pattern and terminates when no better loading pattern can be determined. The process was applied with both reactivity maximization and radial power-peaking minimization as objectives. Results on a typical large PWR indicate that the cost of obtaining an 8% improvement in radial power-peaking margin is ≅2% in fuel cycle costs, for the reload core loaded without burnable poisons that was studied

Core dynamics analysis for reactivity insertion and loss of coolant flow tests using the HTTR

International Nuclear Information System (INIS)

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

2007-01-01

The High Temperature engineering Test Reactor (HTTR) is a graphite-moderated and a gas-cooled reactor with a thermal power of 30 MW and a reactor outlet coolant temperature of 950degC (SAITO, 1994). Safety demonstration tests using the HTTR are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-Temperature Gas-cooled Reactors (HTGRs) (TACHIBANA 2002) (NAKAGAWA 2004). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named ACCORD (TAKAMATSU 2006), was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We used a conventional method, namely, a one-dimensional flow channel model and reactor kinetics model with a single temperature coefficient, taking into account the temperature changes in the core. However, a slight difference between the analytical and experimental results was observed. Therefore, we have modified this code to use a model with four parallel channels and twenty temperature coefficients in the core. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results of the reactivity insertion test as well as the loss of coolant flow tests by tripping one or two out of three gas circulators. Finally, the pre-analytical result of

The effects of applying silicon carbide coating on core reactivity of pebble-bed HTR in water ingress accident

Energy Technology Data Exchange (ETDEWEB)

Zuhair, S.; Setiadipura, Topan [National Nuclear Energy Agency of Indonesia, Serpong Tagerang Selatan (Indonesia). Center for Nuclear Reactor Technology and Safety; Su' ud, Zaki [Bandung Institute of Technology (Indonesia). Dept. of Physics

2017-03-15

Graphite is used as the moderator, fuel barrier material, and core structure in High Temperature Reactors (HTRs). However, despite its good thermal and mechanical properties below the radiation and high temperatures, it cannot avoid corrosion as a consequence of an accident of water/air ingress. Degradation of graphite as a main HTR material and the formation of dangerous CO gas is a serious problem in HTR safety. One of the several steps that can be adopted to avoid or prevent the corrosion of graphite by the water/air ingress is the application of a thin layer of silicon carbide (SiC) on the surface of the fuel element. This study investigates the effect of applying SiC coating on the fuel surfaces of pebble-bed HTR in water ingress accident from the reactivity points of view. A series of reactivity calculations were done with the Monte Carlo transport code MCNPX and continuous energy nuclear data library ENDF/B-VII at temperature of 1200 K. Three options of UO{sub 2}, PuO{sub 2}, and ThO{sub 2}/UO{sub 2} fuel kernel were considered to obtain the inter comparison of the core reactivity of pebble-bed HTR in conditions of water/air ingress accident. The calculation results indicated that the UO{sub 2}-fueled pebble-bed HTR reactivity was slightly reduced and relatively more decreased when the thickness of the SiC coating increased. The reactivity characteristic of ThO{sub 2}/UO{sub 2}-fueled pebble-bed HTR showed a similar trend to that of UO{sub 2}, but did not show reactivity peak caused by water ingress. In contrast with UO{sub 2}- and ThO{sub 2}-fueled pebble-bed HTR, although the reactivity of PuO{sub 2}-fueled pebble-bed HTR was the lowest, its characteristics showed a very high reactivity peak (0.33 Δk/k) and this introduction of positive reactivity is difficult to control. SiC coating on the surface of the plutonium fuel pebble has no significant impact. From the comparison between reactivity characteristics of uranium, thorium and plutonium cores with 0

Study on Characteristic of Temperature Coefficient of Reactivity for Plutonium Core of Pebbled Bed Reactor

Science.gov (United States)

Zuhair; Suwoto; Setiadipura, T.; Bakhri, S.; Sunaryo, G. R.

2018-02-01

As a part of the solution searching for possibility to control the plutonium, a current effort is focused on mechanisms to maximize consumption of plutonium. Plutonium core solution is a unique case in the high temperature reactor which is intended to reduce the accumulation of plutonium. However, the safety performance of the plutonium core which tends to produce a positive temperature coefficient of reactivity should be examined. The pebble bed inherent safety features which are characterized by a negative temperature coefficient of reactivity must be maintained under any circumstances. The purpose of this study is to investigate the characteristic of temperature coefficient of reactivity for plutonium core of pebble bed reactor. A series of calculations with plutonium loading varied from 0.5 g to 1.5 g per fuel pebble were performed by the MCNPX code and ENDF/B-VII library. The calculation results show that the k eff curve of 0.5 g Pu/pebble declines sharply with the increase in fuel burnup while the greater Pu loading per pebble yields k eff curve declines slighter. The fuel with high Pu content per pebble may reach long burnup cycle. From the temperature coefficient point of view, it is concluded that the reactor containing 0.5 g-1.25 g Pu/pebble at high burnup has less favorable safety features if it is operated at high temperature. The use of fuel with Pu content of 1.5 g/pebble at high burnup should be considered carefully from core safety aspect because it could affect transient behavior into a fatal accident situation.

Analysis Influence of Mixing Gd2O3 in the Silicide Fuel Element to Core Excess Reactivity of RSG-GAS

International Nuclear Information System (INIS)

Susilo, Jati

2004-01-01

Gadolinium (Gd 2 O 3 ) is a burnable poison material mixed in the pin fuel element of the LWR core used to decrease core excess reactivity. In this research, analysis influence of mixing Gd 2 O 3 in the silicide fuel element to excess reactivity of the RSG-GAS core had been done. Equivalent cell of the equilibrium core developed by L.E.Strawbridge from Westing House Co. burn-up calculation has been done using SRAC-PIJ computer code achieve infinite multiplication factor (k x ). Value of Gd 2 O 3 concentration in the fuel element (pcm) showed by mass ratio of Gd 2 O 3 (gram) to that U 3 Si 2 (gram) times 10 5 , that is 0 pcm ∼ 100 pcm. From the calculation results analysis showed that Gd 2 O 3 concentration added should be considered. because a large number of Gd 2 O 3 will result in not achieving criticality at the Beginning Of Cycle. The maximum concentration of Gd 2 O 3 for RSG-GAS equilibrium fueled silicide 2.96 grU/cc is 80 pcm or 52.02 mgram/fuel plate. Maximum reduction of core excess reactivity due to mixing of Gd 2 O 3 in the RSG-GAS silicide fuels was around 1.502 %Δk/k, and hence not achieving the standard nominal excess reactivity for RSG-GAS core using high density of U 3 Si 2 -Al fuel. (author)

Uncertainty Evaluation of Reactivity Coefficients for a large advanced SFR Core Design

International Nuclear Information System (INIS)

Khamakhem, Wassim; Rimpault, Gerald

2008-01-01

Sodium Cooled Fast Reactors are currently being reshaped in order to meet Generation IV goals on economics, safety and reliability, sustainability and proliferation resistance. Recent studies have led to large SFR cores for a 3600 MWth power plants, cores which exhibit interesting features. The designs have had to balance between competing aspects such as sustainability and safety characteristics. Sustainability in neutronic terms is translated into positive breeding gain and safety into rather low Na void reactivity effects. The studies have been done on two SFR concepts using oxide and carbide fuels. The use of the sensitivity theory in the ERANOS determinist code system has been used. Calculations have been performed with different sodium evaluations: JEF2.2, ERALIB-1 and the most recent JEFF3.1 and ENDF/B-VII in order to make a broad comparison. Values for the Na void reactivity effect exhibit differences as large as 14% when using the different sodium libraries. Uncertainties due to nuclear data on the reactivity coefficients were performed with BOLNA variances-covariances data, the Na Void Effect uncertainties are near to 12% at 1σ. Since, the uncertainties are far beyond the target accuracy for a design achieving high performance, two directions are envisaged: the first one is to perform new differential measurements or in a second attempt use integral experiments to improve effectively the nuclear data set and its uncertainties such as performed in the past with ERALIB1. (authors)

Reactor core and control rod assembly in FBR type reactor

International Nuclear Information System (INIS)

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

1993-01-01

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

Neutronic characterization of cylindrical core of minor excess reactivity in the nuclear reactor IPEN/MB-01 from the measure of neutron flux distribution and its reactivity ratio

Energy Technology Data Exchange (ETDEWEB)

Bitelli, Ulysses d' Utra; Aredes, Vitor O.G.; Mura, Luiz E.C.; Santos, Diogo F. dos; Silva, Alexandre P. da, E-mail: ubitelli@ipen.br, E-mail: vitoraredes@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2013-07-01

When compared to a rectangular parallelepiped configuration the cylindrical configuration of a nuclear reactor core has a better neutron economy because in this configuration the probability of the neutron leakage is smaller, causing an increase in overall reactivity in the system to the same amount of fuel used. In this work we obtained a critical cylindrical configuration with the control rods 89.50% withdraw from the active region of the IPEN/MB-01 core. This is the cylindrical configuration minimum possible excess of reactivity. Thus we obtained a cylindrical configuration with a diameter of only 28 fuel rods with lowest possible excess of reactivity. For this purpose, 112 peripheral fuel rods are removed from standard reactor core (rectangular parallelepiped of 28x28 fuel rods). In this configuration the excesses of reactivity is approximated 279 pcm. From there, we characterize the neutron field by measuring the spatial distribution of the thermal and epithermal neutron flux for the reactor operating power of 83 watts measured by neutron noise analysis technique and 92.08± 0.07 watts measured by activation technique [10]. The values of thermal and epithermal neutron flux in different directions, axial, radial north-south and radial east-west, are obtained in the asymptotic region of the reactor core, away from the disturbances caused by the reflector and control bar, by irradiating thin gold foils infinitely diluted (1% Au - 99% Al) with and without (bare) cadmium cover. In addition to the distribution of neutron flux, the moderator temperature coefficient, the void coefficient, calibration of the control rods were measured. (author)

Synthesis and Characterization of Core-Shell Acrylate Based Latex and Study of Its Reactive Blends

Directory of Open Access Journals (Sweden)

Ying Nie

2008-03-01

Full Text Available Techniques in resin blending are simple and efficient method for improving the properties of polymers, and have been used widely in polymer modification field. However, polymer latex blends such as the combination of latexes, especially the latexes with water-soluble polymers, were rarely reported. Here, we report a core-shell composite latex synthesized using methyl methacrylate (MMA, butyl acrylate (BA, 2-ethylhexyl acrylate (EHA and glycidyl methacrylate (GMA as monomers and ammonium persulfate and sodium bisulfite redox system as the initiator. Two stages seeded semi-continuous emulsion polymerization were employed for constructing a core-shell structure with P(MMA-co-BA component as the core and P(EHA-co-GMA component as the shell. Results of Transmission Electron Microscopy (TEM and Dynamics Light Scattering (DLS tests confirmed that the particles obtained are indeed possessing a desired core-shell structural character. Stable reactive latex blends were prepared by adding the latex with waterborne melamine-formaldehyde resin (MF or urea-formaldehyde resin (UF. It was found that the glass transition temperature, the mechanical strength and the hygroscopic property of films cast from the latex blends present marked enhancements under higher thermal treatment temperature. It was revealed that the physical properties of chemically reactive latexes with core-shell structure could be altered via the change of crosslinking density both from the addition of crosslinkers and the thermal treatment.

Verification for excess reactivity on beginning equilibrium core of RSG GAS

International Nuclear Information System (INIS)

Daddy Setyawan; Budi Rohman

2011-01-01

BAPETEN is an institution authorized to control the use of nuclear energy in Indonesia. Control for the use of nuclear energy is carried out through three pillars: regulation, licensing, and inspection. In order to assure the safety of the operating research reactors, the assessment unit of BAPETEN is carrying out independent assessment in order to verify safety related parameters in the SAR including neutronic aspect. The work includes verification to the Power Peaking Factor in the equilibrium silicide core of RSG GAS reactor by computational method using MCNP-ORIGEN. This verification calculation results for is 9.4 %. Meanwhile, the RSG-GAS safety analysis report shows that the excess reactivity on equilibrium core of RSG GAS is 9.7 %. The verification calculation results show a good agreement with the report. (author)

Calculation of Reactivity Build up in KANUPP core in Case of Large Break LOCA

International Nuclear Information System (INIS)

Arshad, M. W.

2012-01-01

Loss of Coolant Accident (LOCA) in a Pressurized Heavy Water Reactor (PHWR) leads to coolant expulsion in a primary heat transport system resulting in depressurization and possible core voiding. This results in deterioration of cooling conditions in reactor channels and increase in power before reactor shutdown, leading to higher fuel temperatures.The objective of this thesis is to couple Thermal Hydraulics Data for finding status of 2288 fuel bundles having unique coolant density along with continuous changing state of coolant. WIMCER and CITCER are used for the core calculation in case of LOCA and Thermal Hydraulic Data is obtained from the Thermal Hydraulic code TUF (two unequal flows). These codes are coupled with each other in C programming. Due to degradation of coolant in case of LOCA, the power and reactivity start increasing. Near to 5 mk of reactivity the moderator dump start and reactor goes shut down. The result obtained from these code is followed the same trend as shown in KFSAR. (author)

Expansion of donor-reactive host T cells in primary graft failure after allogeneic hematopoietic SCT following reduced-intensity conditioning.

Science.gov (United States)

Koyama, M; Hashimoto, D; Nagafuji, K; Eto, T; Ohno, Y; Aoyama, K; Iwasaki, H; Miyamoto, T; Hill, G R; Akashi, K; Teshima, T

2014-01-01

Graft rejection remains a major obstacle in allogeneic hematopoietic SCT following reduced-intensity conditioning (RIC-SCT), particularly after cord blood transplantation (CBT). In a murine MHC-mismatched model of RIC-SCT, primary graft rejection was associated with activation and expansion of donor-reactive host T cells in peripheral blood and BM early after SCT. Donor-derived dendritic cells are at least partly involved in host T-cell activation. We then evaluated if such an expansion of host T cells could be associated with graft rejection after RIC-CBT. Expansion of residual host lymphocytes was observed in 4/7 patients with graft rejection at 3 weeks after CBT, but in none of the 17 patients who achieved engraftment. These results suggest the crucial role of residual host T cells after RIC-SCT in graft rejection and expansion of host T cells could be a marker of graft rejection. Development of more efficient T cell-suppressive conditioning regimens may be necessary in the context of RIC-SCT.

Dissecting cross-reactivity in hymenoptera venom allergy by circumvention of alpha-1,3-core fucosylation.

Science.gov (United States)

Seismann, Henning; Blank, Simon; Braren, Ingke; Greunke, Kerstin; Cifuentes, Liliana; Grunwald, Thomas; Bredehorst, Reinhard; Ollert, Markus; Spillner, Edzard

2010-01-01

Hymenoptera venom allergy is known to cause life-threatening and sometimes fatal IgE-mediated anaphylactic reactions in allergic individuals. About 30-50% of patients with insect venom allergy have IgE antibodies that react with both honeybee and yellow jacket venom. Apart from true double sensitisation, IgE against cross-reactive carbohydrate determinants (CCD) are the most frequent cause of multiple reactivities severely hampering the diagnosis and design of therapeutic strategies by clinically irrelevant test results. In this study we addressed allergenic cross-reactivity using a recombinant approach by employing cell lines with variant capacities of alpha-1,3-core fucosylation. The venom hyaluronidases, supposed major allergens implicated in cross-reactivity phenomena, from honeybee (Api m 2) and yellow jacket (Ves v 2a and its putative isoform Ves v 2b) as well as the human alpha-2HS-glycoprotein as control, were produced in different insect cell lines. In stark contrast to production in Trichoplusia ni (HighFive) cells, alpha-1,3-core fucosylation was absent or immunologically negligible after production in Spodoptera frugiperda (Sf9) cells. Consistently, co-expression of honeybee alpha-1,3-fucosyltransferase in Sf9 cells resulted in the reconstitution of CCD reactivity. Re-evaluation of differentially fucosylated hyaluronidases by screening of individual venom-sensitised sera emphasised the allergenic relevance of Api m 2 beyond its carbohydrate epitopes. In contrast, the vespid hyaluronidases, for which a predominance of Ves v 2b could be shown, exhibited pronounced and primary carbohydrate reactivity rendering their relevance in the context of allergy questionable. These findings show that the use of recombinant molecules devoid of CCDs represents a novel strategy with major implications for diagnostic and therapeutic approaches. Copyright 2010 Elsevier Ltd. All rights reserved.

Reactor-core-reactivity control device

International Nuclear Information System (INIS)

Miura, Teruo; Sakuranaga, Tomonobu.

1983-01-01

Purpose: To improve the reactor safety upon failures of control rod drives by adapting a control rod not to drop out accidentally from the reactor core but be inserted into the reactor core. Constitution: The control rod is entered or extracted as usual from the bottom of the pressure vessel. A space is provided above the reactor core within the pressure vessel, in which the moving scope of the control rod is set between the space above the reactor core and the reactor core. That is, the control rod is situated above the reactor core upon extraction thereof and, if an accident occurs to the control rod drive mechanisms to detach the control rod and the driving rod, the control rod falls gravitationally into the reactor core to improve the reactor safety. In addition, since the speed limiter is no more required to the control rod, the driving force can be decreased to reduce the size of the rod drive mechanisms. (Ikeda, J.)

Neutronics aspects associated to irregular lattices in sodium fast reactors cores

International Nuclear Information System (INIS)

Gentili, Michele

2015-01-01

The fuel assemblies of SFR cores (sodium fast reactors) are normally arranged in hexagonal regular lattices, whose compactness is ensured in nominal operating conditions by thermal expansion of assemblies pads disposed on the six assembly wrapper faces. During the reactor operations, thermal expansion phenomena and irradiation creep phenomena occur and they cause the fuel assemblies to bow and to deform both radially and axially. The main goal of this PhD is the understanding of the neutronic aspects and phenomena occurring in case of core and lattice deformations, as much as the design and implementation of deterministic neutronic calculation schemes and methods in order to evaluate the consequences for the core design activities and the safety analysis. The first part of this work is focused on the development of an analytical model with the purpose to identify the neutronic phenomena that are the main contributors to the reactivity changes induced by lattice and core deformations. A first scheme based on the spatial mesh projection method has been conceived and implemented for the ERANOS codes (BISTRO, H3D and VARIANT) and to the SNATCH solver. The second calculation scheme propose is based on mesh deformation: the computing mesh is deformed as a function of the assembly displacement field. This methodology has been implemented for the solver SNATCH, which normally allows the Boltzmann equation to be solved for a regular mesh. Finally, an iterative method has been developed in order to fulfill an a-priori estimation of the maximal reactivity insertion as a function of the postulated mechanical energy provided to the core, as much as the deformation causing it. (author) [fr

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)

A comparative design study of PB-BI cooled reactor cores with forced and natural convection cooling

International Nuclear Information System (INIS)

Mizuno, Tomoyasu; Enuma, Yasuhiro; Tanji, Mikio

2003-01-01

A comparative core design study is performed on Pb-Bi cooled reactors with forced and natural convection (FC and NC) cooling. Major interests of the study are core performance and core safety features. The designed core concepts with nitride fuel achieve reasonable breeding capability. The results of unprotected event analyses such as UTOP and ULOF show that both of concepts have possible features to withstand unprotected events due to negative reactivity feedback by Doppler effect, control rod drive line expansion, etc. These results lead to a conclusion that both of concepts have possible capability as one of future promising core concepts. A FC cooling core concept has more advantage if fuel recycle viewpoint is emphasized. (author)

Reactivity And Neutron Flux At Silicide Fuel Element In The Core Of RSG-GAS

International Nuclear Information System (INIS)

Hamzah, Amir

2000-01-01

In order to 4.8 and 5.2 gr U/cm exp 3 loading of U 3 Si 2 --Al fuel plates characterization, he core reactivity change and neutron flux depression had been done. Control rod calibration method was used to reactivity change measurement and neutron flux distribution was measured using foil activation method. Measurement of insertion of A-type of testing fuel element with U-loading above cannot be done due to technical reason, so the measurement using full type silicide fuel element of 2.96 gr U/cm exp 3 loading. The reactivity change measurement result of insertion in A-9 and C-3 is + 2.67 cent. The flux depression at silicide fuel in A-9 is 1.69 times bigger than oxide and in C-3 is 0.68 times lower than oxide

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

International Nuclear Information System (INIS)

Gimenez, Marcelo; Vertullo, Alicia; Schlamp, Miguel

2000-01-01

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

The Accident Analysis Due to Reactivity Insertion of RSG GAS 3.55 g U/cc Silicide Core

International Nuclear Information System (INIS)

Endiah Puji-Hastuti; Surbakti, Tukiran

2004-01-01

The fuels of RSG-GAS reactor was changed from uranium oxide with 250 g U of loading or 2.96 g U/cc of fuel loading to uranium silicide with the same loading. The silicide fuels can be used in higher density, staying longer in the reactor core and hence having a longer cycle length. The silicide fuel in RSG-GAS core was made up in step-wise by using mixed up core Firstly, it was used silicide fuel with 250 g U of loading and then, silicide fuel with 300 g U of loading (3.55 g U/cc of fuel loading). In every step-wise of fuel loading, it must be analyzed its safety margin. In this occasion, the reactivity accident of RSG-GAS core with 300 g U of silicide fuel loading is analyzed. The calculation was done using EUREKA-2/RR code available at P2TRR. The calculation was done by reactivity insertion at start up and power rangers. The worst case accident is transient due to control rod with drawl failure at start up by means of lowest initial power (0.1 W), either in power range. From all cases which have been done, the results of analysis showed that there is no anomaly and safety margin break at RSG-GAS core with 300 g U silicide fuel loading. (author)

Numerical analysis of the reactivity for the dry lattices above the water level of the critical fuel cores

International Nuclear Information System (INIS)

Nauchi, Yasushi; Kameyama, Takanori

2003-01-01

Criticality analysis has been performed for dozens of tank type cores in which fuel lattices are loaded vertically and partially immersed in light water. The reactivity effect of dry part of lattices stuck above the critical water level has been calculated using the continuous energy Monte Carlo method. The reactivity effect exceeds 0.8% both for MOX and UOX fuel lattices of large buckling (B z 2 > 0.0025 cm -2 ). It is evaluated that at least 20 cm length of fuel rods above the critical water level has significant reactivity effect. (author)

Reactor core

International Nuclear Information System (INIS)

Azekura, Kazuo; Kurihara, Kunitoshi.

1992-01-01

In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

Assessment of CRBR core disruptive accident energetics

International Nuclear Information System (INIS)

Theofanous, T.G.; Bell, C.R.

1984-03-01

The results of an independent assessment of core disruptive accident energetics for the Clinch River Breeder Reactor are presented in this document. This assessment was performed for the Nuclear Regulatory Commission under the direction of the CRBR Program Office within the Office of Nuclear Reactor Regulation. It considered in detail the accident behavior for three accident initiators that are representative of three different classes of events; unprotected loss of flow, unprotected reactivity insertion, and protected loss of heat sink. The primary system's energetics accommodation capability was realistically, yet conservatively, determined in terms of core events. This accommodation capability was found to be equivalent to an isentropic work potential for expansion to one atmosphere of 2550 MJ or a ramp rate of about 200 $/s applied to a classical two-phase disassembly

Insertion of reactivity (RIA) without scram in the reactor core IEA-R1 using code PARET

International Nuclear Information System (INIS)

Alves, Urias F.; Castrillo, Lazara S.; Lima, Fernando A.

2013-01-01

The modeling and analysis thermo hydraulics of a research reactor with MTR type fuel elements - Material Testing Reactor - was performed using the code PARET (Program for the Analysis of Reactor Transients) when in the system some external event is introduced that changed the reactivity in the reactor core. Transients of Reactivity Insertion of 0.5 , 1.5 and 2.0$/ 0.7s in the brazilian reactor IEA-R1 will be presented, and will be shown under what conditions it is possible to ensure the safe operation of its nucleus. (author)

Non-linear triangle-based polynomial expansion nodal method for hexagonal core analysis

International Nuclear Information System (INIS)

Cho, Jin Young; Cho, Byung Oh; Joo, Han Gyu; Zee, Sung Qunn; Park, Sang Yong

2000-09-01

This report is for the implementation of triangle-based polynomial expansion nodal (TPEN) method to MASTER code in conjunction with the coarse mesh finite difference(CMFD) framework for hexagonal core design and analysis. The TPEN method is a variation of the higher order polynomial expansion nodal (HOPEN) method that solves the multi-group neutron diffusion equation in the hexagonal-z geometry. In contrast with the HOPEN method, only two-dimensional intranodal expansion is considered in the TPEN method for a triangular domain. The axial dependence of the intranodal flux is incorporated separately here and it is determined by the nodal expansion method (NEM) for a hexagonal node. For the consistency of node geometry of the MASTER code which is based on hexagon, TPEN solver is coded to solve one hexagonal node which is composed of 6 triangular nodes directly with Gauss elimination scheme. To solve the CMFD linear system efficiently, stabilized bi-conjugate gradient(BiCG) algorithm and Wielandt eigenvalue shift method are adopted. And for the construction of the efficient preconditioner of BiCG algorithm, the incomplete LU(ILU) factorization scheme which has been widely used in two-dimensional problems is used. To apply the ILU factorization scheme to three-dimensional problem, a symmetric Gauss-Seidel Factorization scheme is used. In order to examine the accuracy of the TPEN solution, several eigenvalue benchmark problems and two transient problems, i.e., a realistic VVER1000 and VVER440 rod ejection benchmark problems, were solved and compared with respective references. The results of eigenvalue benchmark problems indicate that non-linear TPEN method is very accurate showing less than 15 pcm of eigenvalue errors and 1% of maximum power errors, and fast enough to solve the three-dimensional VVER-440 problem within 5 seconds on 733MHz PENTIUM-III. In the case of the transient problems, the non-linear TPEN method also shows good results within a few minute of

Reactivity management and burn-up management on JRR-3 silicide-fuel-core

International Nuclear Information System (INIS)

Kato, Tomoaki; Araki, Masaaki; Izumo, Hironobu; Kinase, Masami; Torii, Yoshiya; Murayama, Yoji

2007-08-01

On the conversion from uranium-aluminum-dispersion-type fuel (aluminide fuel) to uranium-silicon-aluminum-dispersion-type fuel (silicide fuel), uranium density was increased from 2.2 to 4.8 g/cm 3 with keeping uranium-235 enrichment of 20%. So, burnable absorbers (cadmium wire) were introduced for decreasing excess reactivity caused by the increasing of uranium density. The burnable absorbers influence reactivity during reactor operation. So, the burning of the burnable absorbers was studied and the influence on reactor operation was made cleared. Furthermore, necessary excess reactivity on beginning of operation cycle and the time limit for restart after unplanned reactor shutdown was calculated. On the conversion, limit of fuel burn-up was increased from 50% to 60%. And the fuel exchange procedure was changed from the six-batch dispersion procedure to the fuel burn-up management procedure. The previous estimation of fuel burn-up was required for the planning of fuel exchange, so that the estimation was carried out by means of past operation data. Finally, a new fuel exchange procedure was proposed for effective use of fuel elements. On the procedure, burn-up of spent fuel was defined for each loading position. The average length of fuel's staying in the core can be increased by two percent on the procedure. (author)

Core design of long life-cycle fast reactors operating without reactivity margin

International Nuclear Information System (INIS)

Aristova, E. N.; Baydin, D. F.; Gol'din, V. Y.; Pestryakova, G. A.; Stoynov, M. I.

2012-01-01

In this paper we consider a possibility of designing a fast reactor core that operates without reactivity margin for a long time. This study is based on the physical principle of fast reactor operating in a self-adjustable neutron-nuclear regime (SANNR-1) introduced by L.P. Feoktistov (1988-1993) and improved by V. Ya. Gol'din SANNR-2 (1995). The mathematical modeling of active zones of fast reactors in SANNR modes is held by authors since 1992. The numerical simulation is based on solving the neutron transport equation coupled with quasi-diffusion equations. The calculations have been performed using standard 26 energy groups. We use a hierarchy of spatial models of 1D, 1.5D, 2D, and 3D geometries. The spatial models of higher dimensionality are used for verification of results. The calculations showed that operation of the reactor in this mode increases its efficiency, safety and simplifies management. It is possible to achieve continuous work of the reactor in SANNR-2 during 7-10 years without fuel overloads by means of further optimization of the mode. Small reactivity margin is used only for the reactor start up. After first 10-15 days the reactor in SANNR-2 operates without reactivity margin. (authors)

Development of a standard data base for FBR core nuclear design (XIII). Analysis of small sample reactivity experiments at ZPPR-9

International Nuclear Information System (INIS)

Sato, Wakaei; Fukushima, Manabu; Ishikawa, Makoto

2000-09-01

A comprehensive study to evaluate and accumulate the abundant results of fast reactor physics is now in progress at O-arai Engineering Center to improve analytical methods and prediction accuracy of nuclear design for large fast breeder cores such as future commercial FBRs. The present report summarizes the analytical results of sample reactivity experiments at ZPPR-9 core, which has not been evaluated by the latest analytical method yet. The intention of the work is to extend and further generalize the standard data base for FBR core nuclear design. The analytical results of the sample reactivity experiments (samples: PU-30, U-6, DU-6, SS-1 and B-1) at ZPPR-9 core in JUPITER series, with the latest nuclear data library JENDL-3.2 and the analytical method which was established by the JUPITER analysis, can be concluded as follows: The region-averaged final C/E values generally agreed with unity within 5% differences at the inner core region. However, the C/E values of every sample showed the radial space-dependency increasing from center to core edge, especially the discrepancy of B-1 was the largest by 10%. Next, the influence of the present analytical results for the ZPPR-9 sample reactivity to the cross-section adjustment was evaluated. The reference case was a unified cross-section set ADJ98 based on the recent JUPITER analysis. As a conclusion, the present analytical results have sufficient physical consistency with other JUPITER data, and possess qualification as a part of the standard data base for FBR nuclear design. (author)

Analysis of fuel options for the breakeven core configuration of the Advanced Recycling Reactor

Energy Technology Data Exchange (ETDEWEB)

Stauff, N.E.; Klim, T.K.; Taiwo, T.A. [Argonne National Laboratory, Argonne, IL (United States); Fiorina, C. [Politecnico di Milano, Milan (Italy); Franceschini, F. [Westinghouse Electric Company LLC., Cranberry Township, Pennsylvania (United States)

2013-07-01

A trade-off study is performed to determine the impacts of various fuel forms on the core design and core physics characteristics of the sodium-cooled Toshiba- Westinghouse Advanced Recycling Reactor (ARR). The fuel forms include oxide, nitride, and metallic forms of U and Th. The ARR core configuration is redesigned with driver and blanket regions in order to achieve breakeven fissile breeding performance with the various fuel types. State-of-the-art core physics tools are used for the analyses. In addition, a quasi-static reactivity balance approach is used for a preliminary comparison of the inherent safety performances of the various fuel options. Thorium-fueled cores exhibit lower breeding ratios and require larger blankets compared to the U-fueled cores, which is detrimental to core compactness and increases reprocessing and manufacturing requirements. The Th cores also exhibit higher reactivity swings through each cycle, which penalizes reactivity control and increases the number of control rods required. On the other hand, using Th leads to drastic reductions in void and coolant expansion coefficients of reactivity, with the potential for enhancing inherent core safety. Among the U-fueled ARR cores, metallic and nitride fuels result in higher breeding ratios due to their higher heavy metal densities. On the other hand, oxide fuels provide a softer spectrum, which increases the Doppler effect and reduces the positive sodium void worth. A lower fuel temperature is obtained with the metallic and nitride fuels due to their higher thermal conductivities and compatibility with sodium bonds. This is especially beneficial from an inherent safety point of view since it facilitates the reactor cool-down during loss of power removal transients. The advantages in terms of inherent safety of nitride and metallic fuels are maintained when using Th fuel. However, there is a lower relative increase in heavy metal density and in breeding ratio going from oxide to metallic

Core 2D. A code for non-isothermal water flow and reactive solute transport. Users manual version 2

Energy Technology Data Exchange (ETDEWEB)

Samper, J.; Juncosa, R.; Delgado, J.; Montenegro, L. [Universidad de A Coruna (Spain)

2000-07-01

Understanding natural groundwater quality patterns, quantifying groundwater pollution and assessing the effects of waste disposal, require modeling tools accounting for water flow, and transport of heat and dissolved species as well as their complex interactions with solid and gases phases. This report contains the users manual of CORE ''2D Version V.2.0, a COde for modeling water flow (saturated and unsaturated), heat transport and multicomponent Reactive solute transport under both local chemical equilibrium and kinetic conditions. it is an updated and improved version of CORE-LE-2D V0 (Samper et al., 1988) which in turns is an extended version of TRANQUI, a previous reactive transport code (ENRESA, 1995). All these codes were developed within the context of Research Projects funded by ENRESA and the European Commission. (Author)

Core2D. A code for non-isothermal water flow and reactive solute transport. Users manual version 2

International Nuclear Information System (INIS)

Samper, J.; Juncosa, R.; Delgado, J.; Montenegro, L.

2000-01-01

Understanding natural groundwater quality patterns, quantifying groundwater pollution and assessing the effects of waste disposal, require modeling tools accounting for water flow, and transport of heat and dissolved species as well as their complex interactions with solid and gases phases. This report contains the users manual of CORE ''2D Version V.2.0, a COde for modeling water flow (saturated and unsaturated), heat transport and multicomponent Reactive solute transport under both local chemical equilibrium and kinetic conditions. it is an updated and improved version of CORE-LE-2D V0 (Samper et al., 1988) which in turns is an extended version of TRANQUI, a previous reactive transport code (ENRESA, 1995). All these codes were developed within the context of Research Projects funded by ENRESA and the European Commission. (Author)

Core 2D. A code for non-isothermal water flow and reactive solute transport. Users manual version 2

Energy Technology Data Exchange (ETDEWEB)

Samper, J; Juncosa, R; Delgado, J; Montenegro, L [Universidad de A Coruna (Spain)

2000-07-01

Understanding natural groundwater quality patterns, quantifying groundwater pollution and assessing the effects of waste disposal, require modeling tools accounting for water flow, and transport of heat and dissolved species as well as their complex interactions with solid and gases phases. This report contains the users manual of CORE ''2D Version V.2.0, a COde for modeling water flow (saturated and unsaturated), heat transport and multicomponent Reactive solute transport under both local chemical equilibrium and kinetic conditions. it is an updated and improved version of CORE-LE-2D V0 (Samper et al., 1988) which in turns is an extended version of TRANQUI, a previous reactive transport code (ENRESA, 1995). All these codes were developed within the context of Research Projects funded by ENRESA and the European Commission. (Author)

Analysis Of Core Management For The Transition Cores Of RSG-GAS Reactor To Full-Silicide Core

International Nuclear Information System (INIS)

Malem Sembiring, Tagor; Suparlina, Lily; Tukiran

2001-01-01

The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 g U/cc is still doing. At the end of 2000, the reactor has been operated for 3 transition cores which is the mixed core of oxide-silicide. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 10 transition cores to achieve a full-silicide core. The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters such as excess reactivity and shutdown margin. The measurement of the core parameters was carried out using the method of compensation of couple control rods. The experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safety to a full-silicide core

The effects of stainless steel radial reflector on core reactivity for small modular reactor

Energy Technology Data Exchange (ETDEWEB)

Kang, Jung Kil, E-mail: jkkang@email.kings.ac.kr; Hah, Chang Joo, E-mail: changhah@kings.ac.kr [KINGS, 658-91, Haemaji-ro, Seosaeng-myeon, Ulju-gun, Ulsan, 689-882 (Korea, Republic of); Cho, Sung Ju, E-mail: sungju@knfc.co.kr; Seong, Ki Bong, E-mail: kbseong@knfc.co.kr [KNFC, Daedeok-daero 989beon-gil, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

2016-01-22

Commercial PWR core is surrounded by a radial reflector, which consists of a baffle and water. Radial reflector is designed to reflect neutron back into the core region to improve the neutron efficiency of the reactor and to protect the reactor vessels from the embrittling effects caused by irradiation during power operation. Reflector also helps to flatten the neutron flux and power distributions in the reactor core. The conceptual nuclear design for boron-free small modular reactor (SMR) under development in Korea requires to have the cycle length of 4∼5 years, rated power of 180 MWth and enrichment less than 5 w/o. The aim of this paper is to analyze the effects of stainless steel radial reflector on the performance of the SMR using UO{sub 2} fuels. Three types of reflectors such as water, water/stainless steel 304 mixture and stainless steel 304 are selected to investigate the effect on core reactivity. Additionally, the thickness of stainless steel and double layer reflector type are also investigated. CASMO-4/SIMULATE-3 code system is used for this analysis. The results of analysis show that single layer stainless steel reflector is the most efficient reflector.

A core design study for 'zero-sodium-void-worth' cores

International Nuclear Information System (INIS)

Kawashima, Masatoshi; Suzuki, Masao; Hill, R.N.

1992-01-01

Recently, a number of low sodium-void-worth metal-fueled core design concepts have been proposed; to provide for flexibility in transuranic nuclide management strategy, core designs which exhibit a wide range of breeding characteristics have been developed. Two core concepts, a flat annular (transuranic burning) core and an absorber-type parfait (transuranic self-sufficient) core, are selected for this study. In this paper, the excess reactivity management schemes applied in the two designs are investigated in detail. In addition, the transient effect of reactivity insertions on the parfait core design is assessed. The upper and lower core regions in the parfait design are neutronically decoupled; however, the common coolant channel creates thermalhydraulic coupling. This combination of neutronic and thermalhydraulic characteristics leads to unique behavior in anticipated transient overpower events. (author)

Method of controlling reactivity

International Nuclear Information System (INIS)

Tochihara, Hiroshi.

1982-01-01

Purpose: To improve the reactivity controlling characteristics by artificially controlling the leakage of neutron from a reactor and providing a controller for controlling the reactivity. Method: A reactor core is divided into several water gaps to increase the leakage of neutron, its reactivity is reduced, a gas-filled control rod or a fuel assembly is inserted into the gap as required, the entire core is coupled in a system to reduce the leakage of the neutron, and the reactivity is increased. The reactor shutdown is conducted by the conventional control rod, and to maintain critical state, boron density varying system is used together. Futher, a control rod drive is used with that similar to the conventional one, thereby enabling fast reactivity variation, and the positive reactivity can be obtained by the insertion, thereby improving the reactivity controlling characteristics. (Yoshihara, H.)

Development and verification of an efficient spatial neutron kinetics method for reactivity-initiated event analyses

International Nuclear Information System (INIS)

Ikeda, Hideaki; Takeda, Toshikazu

2001-01-01

A space/time nodal diffusion code based on the nodal expansion method (NEM), EPISODE, was developed in order to evaluate transient neutron behavior in light water reactor cores. The present code employs the improved quasistatic (IQS) method for spatial neutron kinetics, and neutron flux distribution is numerically obtained by solving the neutron diffusion equation with the nonlinear iteration scheme to achieve fast computation. A predictor-corrector (PC) method developed in the present study enabled to apply a coarse time mesh to the transient spatial neutron calculation than that applicable in the conventional IQS model, which improved computational efficiency further. Its computational advantage was demonstrated by applying to the numerical benchmark problems that simulate reactivity-initiated events, showing reduction of computational times up to a factor of three than the conventional IQS. The thermohydraulics model was also incorporated in EPISODE, and the capability of realistic reactivity event analyses was verified using the SPERT-III/E-Core experimental data. (author)

Reactive Desorption of CO Hydrogenation Products under Cold Pre-stellar Core Conditions

Science.gov (United States)

Chuang, K.-J.; Fedoseev, G.; Qasim, D.; Ioppolo, S.; van Dishoeck, E. F.; Linnartz, H.

2018-02-01

The astronomical gas-phase detection of simple species and small organic molecules in cold pre-stellar cores, with abundances as high as ∼10‑8–10‑9 n H, contradicts the generally accepted idea that at 10 K, such species should be fully frozen out on grain surfaces. A physical or chemical mechanism that results in a net transfer from solid-state species into the gas phase offers a possible explanation. Reactive desorption, i.e., desorption following the exothermic formation of a species, is one of the options that has been proposed. In astronomical models, the fraction of molecules desorbed through this process is handled as a free parameter, as experimental studies quantifying the impact of exothermicity on desorption efficiencies are largely lacking. In this work, we present a detailed laboratory study with the goal of deriving an upper limit for the reactive desorption efficiency of species involved in the CO–H2CO–CH3OH solid-state hydrogenation reaction chain. The limit for the overall reactive desorption fraction is derived by precisely investigating the solid-state elemental carbon budget, using reflection absorption infrared spectroscopy and the calibrated solid-state band-strength values for CO, H2CO and CH3OH. We find that for temperatures in the range of 10 to 14 K, an upper limit of 0.24 ± 0.02 for the overall elemental carbon loss upon CO conversion into CH3OH. This corresponds with an effective reaction desorption fraction of ≤0.07 per hydrogenation step, or ≤0.02 per H-atom induced reaction, assuming that H-atom addition and abstraction reactions equally contribute to the overall reactive desorption fraction along the hydrogenation sequence. The astronomical relevance of this finding is discussed.

Modelling The Effects of Aggregate Size on Alkali Aggregate Reaction Expansion

Directory of Open Access Journals (Sweden)

N. Z. Sekrane

2014-06-01

Full Text Available This work aims at developing models to predict the potential expansion of concrete containing alkali-reactive aggregates. The paper gives measurements in order to provide experimental data concerning the effect of particle size of an alkali-reactive siliceous limestone on mortar expansion. Results show that no expansion was measured on the mortars using small particles (0.5-1.0 mm while the particles (1.0–2.0 mm gave the largest expansions (0.217%. Two models are proposed, the first one studies the correlations between the measured expansions and the size of aggregates, the second one calculates the thickness of the porous zone necessary to take again all the volume of the gel created.

Evaluation of In-Core Fuel Management for the Transition Cores of RSG-GAS Reactor to Full-Silicide Core

International Nuclear Information System (INIS)

S, Tukiran; MS, Tagor; P, Surian

2003-01-01

The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 gU/cc has been done. The core-of RSG-GAS reactor has been operated full core of silicide fuels which is started with the mixed core of oxide-silicide start from core 36. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 9 transition cores (core 36 - 44) to achieve a full-silicide core (core 45). The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters. Conversion core was achieved by transition cores mixed oxide-silicide fuels. Each transition core is calculated and measured core parameter such as, excess reactivity and shutdown margin. Calculation done by Batan-EQUIL-2D code and measurement of the core parameters was carried out using the method of compensation of couple control rods. The results of calculation and experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safely to a full-silicide core

Measurements of the isothermal temperature reactivity coefficient of KUCA C-Core with a D{sub 2}O tank

Energy Technology Data Exchange (ETDEWEB)

Pyeon, Cheol Ho [Research Reactor Institute, Kyoto Univ., Osaka (Japan); Shim, Hyung Jin; Choi, Sung Hoon; Jeon, Byoung Kyu [Seoul National Univ., Seoul (Korea, Republic of); Ryu, Eun Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-10-15

The Kyoto University Critical Assembly (KUCA) is a multi-core type critical assembly consisting of three independent cores in the Kyoto University Research Reactor Institute. The light-water-moderated core (Ccore) is a tank type reactor, and the experiments of the isothermal temperature reactivity coefficient (ITRC) of C-core with a D{sub 2}O tank were carried out with the use of six 10 kW heaters and a radiator system in a dump tank, one 10 kW heater in a core tank, and one 5 kW heater in the D{sub 2}O tank. The ITRCs of the C-core with the D{sub 2}O tank immersed in the core tank are considered important to investigate the mechanism of moderation and reflection effects of H{sub 2}O and D{sub 2}O in the core on the evaluation by numerical simulations. The objectives of this paper are to report the ITRC measurements for C-core with D{sub 2}O tank ranging between 26.7 .deg. C and 58.5 .deg. C, and to examine the accuracy of the numerical simulations by the Seoul National University Monte Carlo code, McCARD, through the comparison between measured and calculated results.

SULEU NTP Core with Passive Reactivity Control and Enhanced Submersion Safety

Energy Technology Data Exchange (ETDEWEB)

Venneri, Paolo; Kim, Yong Hee [KAIST, Daejeon (Korea, Republic of); Eades, Michael J [The Ohio State University, Ohio (United States)

2016-05-15

In this summary, SULEU has been adapted to implement some of the latest developments of LEUNTP design efforts. These include the implementation of a rapid depletion burnable absorber to flatten the reactivity profile during operation and the addition of a lower axial reflector to help minimize the reactivity increase during the full submersion criticality accident. The purpose of this study is to show the state of current LEU-NTP designs in terms of resolving key issues such as minimizing control drum usage and resolving the full submersion criticality accident. Future work will include integrating the rapid depletion poison with other passive reactivity control devices (such as hydrogen density in the tie-tubes) and developing additional systems for mitigating the full submersion criticality accident. It is widely acknowledged that nuclear thermal propulsion (NTP) is an enabling technology for manned missions to Mars and other locations beyond low-Earth orbit. Without nuclear thermal propulsion, manned space travel will be severely limited by the propellant requirements of chemical propulsion and significantly longer travel times of electric propulsion. While the performance superiority of NTP is clear, its implementation has been to date unsuccessful due to the significant costs of development, implementation, and regulations associated with the heritage NTP designs. These new systems take heritage designs and experimental results and adapt them to use LEU fuel with minimum impact on the heritage system. This is done in order to ensure their continued relevance with existing NTP research efforts and enable their rapid implementation into existing NASA efforts for human Mars mission planning. Of the current baseline NTP designs being studied, this paper concerns itself with the improvement of the Superb Use of Low Enriched (SULEU) core.

FBR type reactor core

International Nuclear Information System (INIS)

Tamiya, Tadashi; Kawashima, Katsuyuki; Fujimura, Koji; Murakami, Tomoko.

1995-01-01

Neutron reflectors are disposed at the periphery of a reactor core fuel region and a blanket region, and a neutron shielding region is disposed at the periphery of them. The neutron reflector has a hollow duct structure having a sealed upper portion, a lower portion opened to cooling water, in which a gas and coolants separately sealed in the inside thereof. A driving pressure of a primary recycling pump is lowered upon reduction of coolant flow rate, then the liquid level of coolants in the neutron reflector is lowered due to imbalance between the driving pressure and a gas pressure, so that coolants having an effect as a reflector are eliminated from the outer circumference of the reactor core. Therefore, the amount of neutrons leaking from the reactor core is increased, and negative reactivity is charged to the reactor core. The negative reactivity of the neutron reflector is made greater than a power compensation reactivity. Since this enables reactor scram by using an inherent performance of the reactor core, the reactor core safety of an LMFBR-type reactor can be improved. (I.N.)

Investigation of reactivity variations of the Isfahan MNSR reactor due to variations in the thickness of the core top beryllium layer using WIMSD and MCNP codes

Directory of Open Access Journals (Sweden)

A Shirani

2010-12-01

Full Text Available In this work, the Isfahan Miniature Neutron Source Reactor (MNSR is first simulated using the WIMSD code, and its fuel burn-up after 7 years of operation ( when the reactor was revived by adding a 1.5 mm thick beryllium shim plate to the top of its core and also after 14 years of operation (total operation time of the reactor is calculated. The reactor is then simulated using the MCNP code, and its reactivity variation due to adding a 1.5 mm thick beryllium shim plate to the top of the reactor core, after 7 years of operation, is calculated. The results show good agreement with the available data collected at the revival time. Exess reactivity of the reactor at present time (after 14 years of operation and after 7 years of the the reactor revival time is also determined both experimentally and by calculation, which show good agreement, and indicate that at the present time there is no need to add any further beryllium shim plate to the top of the reactor core. Furthermore, by adding more beryllium layers with various thicknesses to the top of the reactor core, in the input program of the MCNP program, reactivity value of these layers is calculated. From these results, one can predict the necessary beryllium thickness needed to reach a desired reactivity in the MNSR reactor.

Spring unit especially intended for a nuclear reactor core

International Nuclear Information System (INIS)

Brown, S.J.; Gorholt, Wilhelm.

1977-01-01

This invention relates to a spring unit or a group of springs bearing up a sprung mass against an unsprung mass. For instance, a gas cooled high temperature nuclear reactor includes a core of relatively complex structure supported inside a casing or vessel forming a shielded cavity enclosing the reactor core. This core can be assembled from a large number of graphite blocks of different sizes and shapes joined together to form a column. The blocks of each column can be fixed together so as to form together a loose side support. Under the effect of thermal expansion and contraction, shrinkage resulting from irradiation, the effects of pressure and the contraction and creep of the reactor vessel, it is not possible to confine all the columns of the reactor core in a cylindrical rigid structure. Further, the working of the nuclear reactor requires that the reactivity monitoring components may be inserted at any time in the reactor core. A standard process consists in mounting this loosely assembled reactor core in a floating manner by keeping it away from the vessel enclosure around it by means of a number of springs fitted between the lateral surfaces of the core unit and the reactor vessel. The core may be considered as a spring supported mass whereas, relatively, the reactor vessel is a mass that is not flexibly supported [fr

Nuclear data propagation with burnup. Impact on SFR reactivity coefficients

International Nuclear Information System (INIS)

Buiron, Laurent; Plisson-Rieunier, Daniele

2017-01-01

thermal sodium expansion coefficient and the Doppler Effect. Using these sensitivities, a global evaluation of impact of the fuel depletion can be quantified for these reactivity effects at core scale for end of cycle state. An illustration is given for a GEN IV SFR industrial core design (SFR V2B). A first glance at preliminary uncertainty level is presented using current covariance matrices available at CEA. (author)

(Gold core)/(titania shell) nanostructures for plasmon-enhanced photon harvesting and generation of reactive oxygen species

KAUST Repository

Fang, Caihong; Jia, Henglei; Chang, Shuai; Ruan, Qifeng; Wang, Peng; Chen, Tao; Wang, Jianfang

2014-01-01

Integration of gold and titania in a nanoscale core/shell architecture can offer large active metal/semiconductor interfacial areas and avoid aggregation and reshaping of the metal nanocrystal core. Such hybrid nanostructures are very useful for studying plasmon-enhanced/enabled processes and have great potential in light-harvesting applications. Herein we report on a facile route to (gold nanocrystal core)/(titania shell) nanostructures with their plasmon band synthetically variable from ∼700 nm to over 1000 nm. The coating method has also been applied to other mono- and bi-metallic Pd, Pt, Au nanocrystals. The gold/titania nanostructures have been employed as the scattering layer in dye-sensitized solar cells, with the resultant cells exhibiting a 13.3% increase in the power conversion efficiency and a 75% decrease in the scattering-layer thickness. Moreover, under resonant excitation, the gold/titania nanostructures can efficiently utilize low-energy photons to generate reactive oxygen species, including singlet oxygen and hydroxyl radicals.

Mineralogy, geochemistry and expansion testing of an alkali-reactive basalt from western Anatolia, Turkey

International Nuclear Information System (INIS)

Copuroglu, Oguzhan; Andic-Cakir, Ozge; Broekmans, Maarten A.T.M.; Kuehnel, Radko

2009-01-01

In this paper, the alkali-silica reaction performance of a basalt rock from western Anatolia, Turkey is reported. It is observed that the rock causes severe gel formation in the concrete microbar test. It appears that the main source of expansion is the reactive glassy phase of the basalt matrix having approximately 70% of SiO 2 . The study presents the microstructural characteristics of unreacted and reacted basalt aggregate by optical and electron microscopy and discusses the possible reaction mechanism. Microstructural analysis revealed that the dissolution of silica is overwhelming in the matrix of the basalt and it eventually generates four consequences: (1) Formation of alkali-silica reaction gel at the aggregate perimeter, (2) increased porosity and permeability of the basalt matrix, (3) reduction of mechanical properties of the aggregate and (4) additional gel formation within the aggregate. It is concluded that the basalt rock is highly prone to alkali-silica reaction. As an aggregate, this rock is not suitable for concrete production.

Mineralogy, geochemistry and expansion testing of an alkali-reactive basalt from western Anatolia, Turkey

Energy Technology Data Exchange (ETDEWEB)

Copuroglu, Oguzhan, E-mail: O.Copuroglu@CiTG.TUDelft.NL [Delft University of Technology, Faculty of CiTG, Materials and Environment, Stevinweg 1, 2628CN, Delft (Netherlands); Andic-Cakir, Ozge [Ege University, Civil Engineering Dept., 35100 Bornova, Izmir (Turkey); Broekmans, Maarten A.T.M. [Geological Survey of Norway, Dept. of Mineral Characterization, N-7491 Trondheim (Norway); Kuehnel, Radko [Burgemeester Merkusstraat 5, 2645 NJ, Delfgauw (Netherlands)

2009-07-15

In this paper, the alkali-silica reaction performance of a basalt rock from western Anatolia, Turkey is reported. It is observed that the rock causes severe gel formation in the concrete microbar test. It appears that the main source of expansion is the reactive glassy phase of the basalt matrix having approximately 70% of SiO{sub 2}. The study presents the microstructural characteristics of unreacted and reacted basalt aggregate by optical and electron microscopy and discusses the possible reaction mechanism. Microstructural analysis revealed that the dissolution of silica is overwhelming in the matrix of the basalt and it eventually generates four consequences: (1) Formation of alkali-silica reaction gel at the aggregate perimeter, (2) increased porosity and permeability of the basalt matrix, (3) reduction of mechanical properties of the aggregate and (4) additional gel formation within the aggregate. It is concluded that the basalt rock is highly prone to alkali-silica reaction. As an aggregate, this rock is not suitable for concrete production.

Stationary liquid fuel fast reactor SLFFR – Part I: Core design

Energy Technology Data Exchange (ETDEWEB)

Jing, T.; Yang, G.; Jung, Y.S.; Yang, W.S., E-mail: yang494@purdue.edu

2016-12-15

Highlights: • An innovative fast reactor concept SLFFR based on liquid metal fuel is proposed for TRU burning. • A compact core design of 1000 MWt SLFFR is developed to achieve a zero conversion ratio and passive safety. • The core size and the control requirement are significantly reduced compared to the conventional solid fuel reactor with same conversion ratio. - Abstract: For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named the stationary liquid fuel fast reactor (SLFFR) has been proposed based on a stationary molten metallic fuel. A compact core design of a 1000 MWt SLFFR has been developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches have been adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses have been performed to evaluate the steady-state performance characteristics. The analysis results indicate that the SLFFR of a zero TRU conversion ratio is feasible while satisfying the conservatively imposed thermal design constraints. A theoretical maximum TRU consumption rate of 1.01 kg/day is achieved with uranium-free fuel. Compared to the solid fuel reactors with the same TRU conversion ratio, the core size and the reactivity control requirement are reduced significantly. The primary and secondary control systems provide sufficient shutdown margins, and the calculated reactivity feedback coefficients show that the prompt fuel expansion coefficient is sufficiently negative.

Virial Expansion Bounds

Science.gov (United States)

Tate, Stephen James

2013-10-01

In the 1960s, the technique of using cluster expansion bounds in order to achieve bounds on the virial expansion was developed by Lebowitz and Penrose (J. Math. Phys. 5:841, 1964) and Ruelle (Statistical Mechanics: Rigorous Results. Benjamin, Elmsford, 1969). This technique is generalised to more recent cluster expansion bounds by Poghosyan and Ueltschi (J. Math. Phys. 50:053509, 2009), which are related to the work of Procacci (J. Stat. Phys. 129:171, 2007) and the tree-graph identity, detailed by Brydges (Phénomènes Critiques, Systèmes Aléatoires, Théories de Jauge. Les Houches 1984, pp. 129-183, 1986). The bounds achieved by Lebowitz and Penrose can also be sharpened by doing the actual optimisation and achieving expressions in terms of the Lambert W-function. The different bound from the cluster expansion shows some improvements for bounds on the convergence of the virial expansion in the case of positive potentials, which are allowed to have a hard core.

Insertion material for controlling reactivity

International Nuclear Information System (INIS)

Baba, Iwao.

1994-01-01

Moderators and a group of suspended materials having substantially the same density as the moderator are sealed in a hollow rod vertically inserted to a fuel assembly. Specifically, the group of suspended materials is adapted to have a density changing stepwise from density of the moderator at the exit temperature of the reactor core to that at the inlet temperature of the reactor core. Reactivity is selectively controlled for a portion of high power and a portion of high reactivity by utilizing the density of the moderator and the distribution of the density. That is, if the power distribution is flat, the density of the moderators changes at a constant rate over the vertical direction of the reactor core and the suspended materials stay at a portion of the same density, to form a uniform distribution. Further, upon reactor shutdown, since the liquid temperature of the moderators is lowered and the density is increased, all of beads are collected at the upper portion to remove water at the upper portion of the reactor core of low burnup degree thereby selectively controlling the reactivity at a portion of high power and a portion of high reactivity. (N.H.)

Automated reactivity anomaly surveillance in the Fast Flux Test Facility

International Nuclear Information System (INIS)

Knutson, B.J.; Harris, R.A.; Honeyman, D.J.; Shook, A.T.; Krohn, C.N.

1985-01-01

The automated technique for monitoring core reactivity during power operation used at the Fast Flux Test Facility (FFTF) is described. This technique relies on comparing predicted to measured rod positions to detect any anomalous (or unpredicted) core reactivity changes. It is implemented on the Plant Data System (PDS) computer and, thus, provides rapid indication of any abnormal core conditions. The prediction algorithms use thermal-hydraulic, control rod position and neutron flux sensor information to predict the core reactivity state

The influence of the reactivity ramp on the course of the power transient in the MARK 1A core of the SNR 300

International Nuclear Information System (INIS)

Froehlich, R.; Schmuck, P.

1976-01-01

The course of a hypothetic transient overpower accident caused by the onset of a not further specified reactivity ramp accompanied by the simultaneous failure of both shutdown systems must be analyzed in the SNR 300 Mark 1A core licensing procedure. The present study is limited to the discussion of the starting and shutdown phases of such accidents for the fresh core. Depending on the operational state of the reactor, the core geometry is still intact during the starting phase. In the following shutdown phase (core disassembly phase), large-scale mass transfer leads to the nuclear shutdown of the reactor. (orig./AK) [de

Development of Regulatory Audit Core Safety Code : COREDAX

Energy Technology Data Exchange (ETDEWEB)

Yang, Chae Yong; Jo, Jong Chull; Roh, Byung Hwan [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Jae Jun; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

2005-07-01

Korea Institute of Nuclear Safety (KINS) has developed a core neutronics simulator, COREDAX code, for verifying core safety of SMART-P reactor, which is technically supported by Korea Advanced Institute of Science and Technology (KAIST). The COREDAX code would be used for regulatory audit calculations of 3- dimendional core neutronics. The COREDAX code solves the steady-state and timedependent multi-group neutron diffusion equation in hexagonal geometry as well as rectangular geometry by analytic function expansion nodal (AFEN) method. AFEN method was developed at KAIST, and it was internationally verified that its accuracy is excellent. The COREDAX code is originally programmed based on the AFEN method. Accuracy of the code on the AFEN method was excellent for the hexagonal 2-dimensional problems, but there was a need for improvement for hexagonal-z 3-dimensional problems. Hence, several solution routines of the AFEN method are improved, and finally the advanced AFEN method is created. COREDAX code is based on the advanced AFEN method . The initial version of COREDAX code is to complete a basic framework, performing eigenvalue calculations and kinetics calculations with thermal-hydraulic feedbacks, for audit calculations of steady-state core design and reactivity-induced accidents of SMART-P reactor. This study describes the COREDAX code for hexagonal geometry.

Core reactivity estimation in space reactors using recurrent dynamic networks

Science.gov (United States)

Parlos, Alexander G.; Tsai, Wei K.

1991-01-01

A recurrent multilayer perceptron network topology is used in the identification of nonlinear dynamic systems from only the input/output measurements. The identification is performed in the discrete time domain, with the learning algorithm being a modified form of the back propagation (BP) rule. The recurrent dynamic network (RDN) developed is applied for the total core reactivity prediction of a spacecraft reactor from only neutronic power level measurements. Results indicate that the RDN can reproduce the nonlinear response of the reactor while keeping the number of nodes roughly equal to the relative order of the system. As accuracy requirements are increased, the number of required nodes also increases, however, the order of the RDN necessary to obtain such results is still in the same order of magnitude as the order of the mathematical model of the system. It is believed that use of the recurrent MLP structure with a variety of different learning algorithms may prove useful in utilizing artificial neural networks for recognition, classification, and prediction of dynamic systems.

A study of temperature coefficients of reactivity for a Savannah River Site tritium-producing charge

International Nuclear Information System (INIS)

George, D.L.; Frost, R.L.

1991-01-01

Temperature coefficients of reactivity have been calculated for the Mark 22 assembly in the K-14 charge at the Savannah River Site. Temperature coefficients are the most important reactivity feedback mechanism in SRS reactors; they are used in all safety analyses performed in support of the Safety Analysis Report, and in operations to predict reactivity changes with control rod moves. The effects of the radial location of the assembly in the reactor, isotope depletion, and thermal expansion of the metal components on the temperature coefficients have also been investigated. With the exception of the dead space coefficient, all of the regional temperature coefficients were found to be negative or zero. All of the temperature coefficients become more negative with isotopic depletion over the fuel cycle. Coefficients also become more negative with increasing radial distance of the assembly from the center of the core; this is proven from first principles and confirmed by calculations. It was found that axial and radial thermal expansion effects on the metal fuel and target tubes counteract one another, indicating these effects do not need to be considered in future temperature coefficient calculations for the Mark 22 assembly. The moderator coefficient was found to be nonlinear with temperature; thus, the values derived for accidents involving increases in moderator temperature are significantly different than those for decreases in moderator temperature, although the moderator coefficient is always negative

Automatic determination of pressurized water reactor core loading patterns which maximize end-of-cycle reactivity within power peaking and burnup constraints

International Nuclear Information System (INIS)

Hobson, G.H.

1985-01-01

An automated procedure for determining the optimal core loading pattern for a pressurized water reactor which maximizes end-of-cycle k/sub eff/ while satisfying constraints on power peaking and discharge burnup has been developed. The optimization algorithm combines a two energy group, two-dimensional coarse-mesh finite difference diffusion theory neutronics model to simulate core conditions, a perturbation theory approach to determine reactivity, flux, power and burnup changes as a function of assembly shuffling, and Monte Carlo integer programming to select the optimal loading pattern solution. The core examined was a typical Cycle 2 reload with no burnable poisons. Results indicate that the core loading pattern that maximizes end-of-cycle k/sub eff/ results in a 5.4% decrease in fuel cycle costs compared with the core loading pattern that minimizes the maximum relative radial power peak

Three-dimensional static and dynamic reactor calculations by the nodal expansion method

International Nuclear Information System (INIS)

Christensen, B.

1985-05-01

This report reviews various method for the calculation of the neutron-flux- and power distribution in an nuclear reactor. The nodal expansion method (NEM) is especially described in much detail. The nodal expansion method solves the diffusion equation. In this method the reactor core is divided into nodes, typically 10 to 20 cm in each direction, and the average flux in each node is calculated. To obtain the coupling between the nodes the local flux inside each node is expressed by use of a polynomial expansion. The expansion is one-dimensional, so inside each node such three expansions occur. To calculate the expansion coefficients it is necessary that the polynomial expansion is a solution to the one-dimensional diffusion equation. When the one-dimensional diffusion equation is established a term with the transversal leakage occur, and this term is expanded after the same polynomials. The resulting equation system with the expansion coefficients as the unknowns is solved with weigthed residual technique. The nodal expansion method is built into a computer program (also called NEM), which is divided into two parts, one part for steady-state calculations and one part for dynamic calculations. It is possible to take advantage of symmetry properties of the reactor core. The program is very flexible with regard to the number of energy groups, the node size, the flux expansion order and the transverse leakage expansion order. The boundary of the core is described by albedos. The program and input to it are described. The program is tested on a number of examples extending from small theoretical one up to realistic reactor cores. Many calculations are done on the wellknown IAEA benchmark case. The calculations have tested the accuracy and the computing time for various node sizes and polynomial expansions. In the dynamic examples various strategies for variation of the time step-length have been tested. (author)

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

MTR fuel element burn-up measurements by the reactivity method

International Nuclear Information System (INIS)

Zuniga, A.; Cuya, T.R.; Ravnik, M.

2003-01-01

Fuel element burn-up was measured by the reactivity method in the 10 MW Peruvian MTR reactor RP-10. The main purpose of the experiment was testing the reactivity method for an MTR reactor as the reactivity method was originally developed for TRIGA reactors. The reactivity worth of each measured fuel element was measured in its original core position in order to measure the burn-up of the fuel elements that were part of the experimental core. The burn-up of each measured fuel element was derived by interpolating its reactivity worth from the reactivity worth of two reference fuel elements of known burn-up, whose reactivity worth was measured in the position of the measured fuel element. The accuracy of the method was improved by separating the reactivity effect of burn-up from the effect of the position in the core. The results of the experiment showed that the modified reactivity method for fuel element burn-up determination could be applied also to MTR reactors. (orig.)

Fort St. Vrain core performance

International Nuclear Information System (INIS)

McEachern, D.W.; Brown, J.R.; Heller, R.A.; Franek, W.J.

1977-07-01

The Fort St. Vrain High Temperature Gas Cooled Reactor core performance has been evaluated during the startup testing phase of the reactor operation. The reactor is graphite moderated, helium cooled, and uses coated particle fuel and on-line flow control to each of the 37 refueling regions. Principal objectives of startup testing were to determine: core and control system reactivity, radial power distribution, flow control capability, and initial fission product release. Information from the core demonstrates that Technical Specifications are being met, performance of the core and fuel is as expected, flow and reactivity control are predictable and simple for the operator to carry out

Removal of Reactive Anionic Dyes from Binary Solutions by Adsorption onto Quaternized Kenaf Core Fiber

Directory of Open Access Journals (Sweden)

Intidhar Jabir Idan

2017-01-01

Full Text Available The most challenging mission in wastewater treatment plants is the removal of anionic dyes, because they are water-soluble and produce very shining colours in the water. In this regard, kenaf core fiber (KCF was chemically modified by the quaternized agent (3-chloro-2-hydroxypropyltrimethylammonium chloride to increase surface area and change the surface properties in order to improve the removing reactive anionic dyes from binary aqueous solution. The influencing operating factors like dye concentration, pH, adsorbent dosage, and contact time were examined in a batch mode. The results indicate that the percentage of removal of Reactive Red-RB (RR-RB and Reactive Black-5 (RB-5 dyes from binary solution was increased with increasing dyes concentrations and the maximum percentage of removal reached up to 98.4% and 99.9% for RR-RB and RB-5, respectively. Studies on effect of pH showed that the adsorption was not significantly influenced by pH. The equilibrium analyses explain that, in spite of the extended Langmuir model failure to describe the data in the binary system, it is better than the Jain and Snoeyink model in describing the adsorption behavior of binary dyes onto QKCF. Also, the pseudo-second-order model was better to represent the adsorption kinetics for RR-RB and RB-5 dyes on QKCF.

Removal of Reactive Orange 16 Dye from Aqueous Solution by Using Modified Kenaf Core Fiber

Directory of Open Access Journals (Sweden)

Maytham Kadhim Obaid

2016-01-01

Full Text Available Evaluated removal of reactive orange 16 (RO16 dye from aqueous solution was studied in batch mode by using kenaf core fiber as low-cost adsorbents. In this attempt, kenaf core fiber with size 0.25–1 mm was treated by using (3-chloro-2-hydroxypropyl trimethylammonium chloride (CHMAC as quaternization agent. Then effective parameters include adsorbent dose, pH, and contact time and initial dye concentration on adsorption by modified kenaf core fiber was investigated. In addition, isotherms and kinetics adsorption studies were estimated for determination of the equilibrium adsorption capacity and reactions dynamics, respectively. Results showed that the best dose of MKCF was 0.1 g/100 mL, the maximum removal of RO16 was 97.25 at 30°C, pH = 6.5, and agitation speed was 150 rpm. The results also showed that the equilibrium data were represented by Freundlich isotherm with correlation coefficients R2=0.9924, and the kinetic study followed the pseudo-second-order kinetic model with correlation coefficients R2=0.9997 for Co=100 mg/L. Furthermore, the maximum adsorption capacity was 416.86 mg/g. Adsorption through kenaf was found to be very effective for the removal of the RO16 dye.

Correlations among FBR core characteristics for various fuel compositions

International Nuclear Information System (INIS)

Maruyama, Shuhei; Ohki, Shigeo; Okubo, Tsutomu; Kawashima, Katsuyuki; Mizuno, Tomoyasu

2012-01-01

In the design of a fast breeder reactor (FBR) core for the light water reactor (LWR) to FBR transition stage, it is indispensable to grasp the effect of a wide range of fuel composition variations on the core characteristics. This study finds good correlations between burnup reactivity and safety parameters, such as the sodium void reactivity and Doppler coefficient, for various fuel compositions and determines the mechanisms behind these correlations with the aid of sensitivity analyses. It is clarified that the Doppler coefficient is actually correlated with the other core characteristics by considering the constraint imposed by the requirement of sustaining criticality on the fuel composition variations. These correlations make it easy to specify the various properties ranges for core reactivity control and core safety, which are important for core design in determining the core specifications and performance. They provide significant information for FBR core design for the transition stage. Moreover, as an application of the above-mentioned correlations, a simplified burnup reactivity index is developed for rapid and rational estimation of the core characteristic variations. With the use of this index and these correlations, the core characteristic variations can be estimated for various fuel compositions without repeating the core calculations. (author)

Evaluation of reactivity and Xe behavior during daily load following operation

International Nuclear Information System (INIS)

Sakamoto, Yasunori; Araki, Tsuneyasu; Yamamoto, Fumiaki

1992-01-01

A boiling water reactor (BWR) has an excellent load following capability provided by a core flow control, which is used for changing a reactor power level and for compensating the subsequent Xe concentration change. The core characteristics during load following operations are investigated in detail, using our reactor core simulator. Comparisons of changes of the Doppler reactivity, the void reactivity and the Xe reactivity during transients are performed. Also the features of Xe transient during load following operations are shown. It has been shown that the core flow change required to compensate the Xe reactivity change produces much greater change of the void reactivity than that required for power level changes, and that the resulting local power change in the lower part of the core is greater than that in the upper part, because the Xe concentration change in the lower part is hardly compensated by the core flow control. Also the effects of power level changes, cycle patterns, and initial concentration of Xe and I on the Xe transient behavior have been investigated. (author)

Conceptual design study of LMFBR core with carbide fuel

International Nuclear Information System (INIS)

Tezuka, H.; Hojuyama, T.; Osada, H.; Ishii, T.; Hattori, S.; Nishimura, T.

1987-01-01

Carbide fuel is a hopeful candidate for demonstration FBR(DFBR) fuel from the plant cost reduction point of view. High thermal conductivity and high heavy metal content of carbide fuel lead to high linear heat rate and high breeding ratio. We have analyzed carbide fuel core characteristics and have clarified the concept of carbide fuel core. By survey calculation, we have obtained a correlation map between core parameters and core characteristics. From the map, we have selected a high efficiency core whose features are better than those of an oxide core, and have obtained reactivity coefficients. The core volume and the reactor fuel inventory are approximately 20% smaller, and the burn-up reactivity loss is 50% smaller compared with the oxide fuel core. These results will reduce the capital cost. The core reactivity coefficients are similar to the conventional oxide DFBR's. Therefore the carbide fuel core is regarded as safe as the oxide core. Except neutron fluence, the carbide fuel core has better nuclear features than the oxide core

Reactor core fuel management

International Nuclear Information System (INIS)

Silvennoinen, P.

1976-01-01

The subject is covered in chapters, entitled: concepts of reactor physics; neutron diffusion; core heat transfer; reactivity; reactor operation; variables of core management; computer code modules; alternative reactor concepts; methods of optimization; general system aspects. (U.K.)

Analysis of Moderator Temperature Reactivity Coefficient of the PWR Core Using WIMS-ANL

International Nuclear Information System (INIS)

Tukiran; Rokhmadi

2007-01-01

The Moderator Temperature Reactivity Coefficient (MTRC) is an important parameter in design, control and safety, particularly in PWR reactor. It is then very important to validate any new processed library for an accurate prediction of this parameter. The objective of this work is to validate the newly WIMS library based on ENDF/B-VI nuclear data files, especially for the prediction of the MTRC parameter. For this purpose, it is used a set of light water moderated lattice experiments as the NORA experiment and R1-100H critical reactors, both of reactors using UO 2 fuel pellet. Analysis is used with WIMSD/4 lattice code with original cross section libraries and WIMS-ANL with ENDF/B-VI cross section libraries. The results showed that the moderator temperatures reactivity coefficients for the NORA reactor using original libraries is - 5.039E-04 %Δk/k/℃ but for ENDF/B-VI libraries is - 2.925E-03 %Δk/k/℃. Compared to the designed value of the reactor core, the difference is in the range of 1.8 - 3.8 % for ENDF/B-IV libraries. It can be concluded that for reactor safety and control analysis, it has to be used ENDF/B- VI libraries because the original libraries is not accurate any more. (author)

Adiabatic supernova expansion into the circumstellar medium

International Nuclear Information System (INIS)

Band, D.L.; Liang, E.P.

1987-01-01

We perform one dimensional numerical simulations with a Lagrangian hydrodynamics code of the adiabatic expansion of a supernova into the surrounding medium. The early expansion follows Chevalier's analytic self-similar solution until the reverse shock reaches the ejecta core. We follow the expansion as it evolves towards the adiabatic blast wave phase. Some memory of the earlier phases of expansion is retained in the interior even when the outer regions expand as a blast wave. We find the results are sensitive to the initial configuration of the ejecta and to the placement of gridpoints. 6 refs., 2 figs

Reactor core in FBR type reactor

International Nuclear Information System (INIS)

Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

1989-01-01

In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

Neutronic characterization of cylindrical core of minor excess reactivity in the nuclear reactor IPEN/MB-01 from the measure of spatial and energetic distribution of neutron flux distribution

International Nuclear Information System (INIS)

Aredes, Vitor Ottoni Garcia

2014-01-01

In this work was conducted the mapping of the thermal and epithermal neutrons flux and the energy spectrum of the neutrons in the reactor core IPEN/MB-01 for a cylindrical core configuration with minor excess reactivity, which is 28 x 28 fuel rods arranged in north-south and east-west directions. The calibration of control rods for this configuration determined their excess reactivity. The lower excess reactivity in the core decreased neutron flux disturbance caused by the neutron absorbing rods , given that the nuclear reactor was operated with the rods almost completely removed . Was used the 'Activation Analysis Technique' with the thin foil activation detectors ( infinitely diluted and hyper-pure), of different materials that work in different energy ranges, to calculate the saturation activity, used for determining the neutron flux and in the SANDBP code as input for the calculation of the neutrons energy spectrum. To discriminate thermal and epithermal flux , was used the 'Cadmium RatioTechnique' . The activation detectors were distributed in a total of 140 radial and axial positions in the reactor core and 16 irradiation, with bare and covered with cadmium activation foils. A model of this configuration was simulated by MCNP-5 code to determine the cadmium correction factor and comparison of the results obtained experimentally. The cylindrical configuration desired, with 17% less fuel than the standard rectangular configuration (28 x 26 fuel rods), reached criticality with the control rods approximately 90% removed, which decreased considerably the disturbance in neutron flux. Given the highest power density of the 28 x 28 cylindrical core, the neutron flux increased by over 50% in the central regions of the core compared to the values of the 28 x 26 standard rectangular core. (author)

Regarding KUR Reactivity Measurement System

International Nuclear Information System (INIS)

Nakamori, Akira; Hasegawa, Kei; Tsuchiyama, Tatsuo; Yamamoto, Toshihiro; Okumura, Ryo; Sano, Tadafumi

2012-01-01

This article reported: (1) the outline of the reactivity measurement system of Kyoto University Research Reactor (KUR), (2) the calibration data of control rod, (3) the problems and the countermeasures for range switching of linear output meter. For the laptop PC for the reactivity measurement system, there are four input signals: (1) linear output meter, (2) logarithmic output meter, (3) core temperature gauge, and (4) control rod position. The hardware of reactivity measurement system is controlled with Labview installed on the laptop. Output, reactivity, reactor period, and the change in reactivity due to temperature effect or Xenon effect are internally calculated and displayed in real-time with Labview based on the four signals above. Calculation results are recorded in the form of a spreadsheet. At KUR, the reactor core arrangement was changed, so the control rod was re-calibrated. At this time, calculated and experimental values of reactivity based on the reactivity measurement system were compared, and it was confirmed that the reactivity calculation by Labview was accurate. The range switching of linear output meter in the nuclear instrumentation should automatically change within the laptop, however sometimes this did not function properly in the early stage. It was speculated that undefined percent values during the transition of percent value were included in the calculation and caused calculation errors. The range switching started working properly after fixing this issue. (S.K.)

In-situ Density and Thermal Expansion Measurements of Fe and Fe-S Alloying Liquids Under Planetary Core Conditions

Science.gov (United States)

Jing, Z.; Chantel, J.; Yu, T.; Sakamaki, T.; Wang, Y.

2015-12-01

Liquid iron is likely the dominant constituent in the cores of terrestrial planets and icy satellites such as Earth, Mars, Mercury, the Moon, Ganymede, and Io. Suggested by geophysical and geochemical observations, light elements such as S, C, Si, etc., are likely present in planetary cores. These light elements can significantly reduce the density and melting temperature of the Fe cores, and hence their abundances are crucial to our understanding of the structure and thermal history of planetary cores, as well as the generation of intrinsic magnetic fields. Knowledge on the density of Fe-light element alloying liquids at high pressures is critical to place constraints on the composition of planetary cores. However, density data on liquid Fe-light element alloys at core pressures are very limited in pressure and composition and are sometimes controversial. In this study, we extend the density dataset for Fe-rich liquids by measuring the density of Fe, Fe-10wt%S, Fe-20wt%S, Fe-27wt%S, and FeS liquids using the X-ray absorption technique in a DIA-type multianvil apparatus up to 7 GPa and 2173 K. An ion chamber (1D-detector) and a CCD camera (2D-detector) were used to measure intensities of transmitted monochromatic X-rays through molten samples, with the photon energy optimized at 40 keV. The densities were then determined from the Beer-Lambert law using the mass absorption coefficients, calibrated by solid standards using X-ray diffraction. At each pressure, density measurements were conducted at a range of temperatures above the liquidus of the samples, enabling the determination of thermal expansion. Combined with our previous results on the sound velocity of Fe and Fe-S liquids at high pressures (Jing et al., 2014, Earth Planet. Sci. Lett. 396, 78-87), these data provide tight constraints on the equation of state and thermodynamic properties such as the adiabatic temperature gradient for Fe-S liquids. We will discuss these results with implications to planetary

Localized reactive flow in carbonate rocks: Core-flood experiments and network simulations

Science.gov (United States)

Wang, Haoyue; Bernabé, Yves; Mok, Ulrich; Evans, Brian

2016-11-01

We conducted four core-flood experiments on samples of a micritic, reef limestone from Abu Dhabi under conditions of constant flow rate. The pore fluid was water in equilibrium with CO2, which, because of its lowered pH, is chemically reactive with the limestone. Flow rates were between 0.03 and 0.1 mL/min. The difference between up and downstream pore pressures dropped to final values ≪1 MPa over periods of 3-18 h. Scanning electron microscope and microtomography imaging of the starting material showed that the limestone is mostly calcite and lacks connected macroporosity and that the prevailing pores are few microns large. During each experiment, a wormhole formed by localized dissolution, an observation consistent with the decreases in pressure head between the up and downstream reservoirs. Moreover, we numerically modeled the changes in permeability during the experiments. We devised a network approach that separated the pore space into competing subnetworks of pipes. Thus, the problem was framed as a competition of flow of the reactive fluid among the adversary subnetworks. The precondition for localization within certain time is that the leading subnetwork rapidly becomes more transmissible than its competitors. This novel model successfully simulated features of the shape of the wormhole as it grew from few to about 100 µm, matched the pressure history patterns, and yielded the correct order of magnitude of the breakthrough time. Finally, we systematically studied the impact of changing the statistical parameters of the subnetworks. Larger mean radius and spatial correlation of the leading subnetwork led to faster localization.

Laboratory measurements of the coefficient of thermal expansion of Olkiluoto drill core samples

International Nuclear Information System (INIS)

Aakesson, U.

2012-04-01

The coefficient of thermal expansion and the wet density has been determined on 22 specimens from the ONKALO drillholes ONK-PP167, ONK-PP199, ONK-PP224, ONK-PP225 and ONK-PP226, Olkiluoto, Finland. The coefficient of thermal expansion has been determined in the temperature interval 20-60 deg C. The results indicated that the thermal expansion was almost linear, and the coefficient of thermal expansion for the investigated specimens range between 3.2 and 14.4 x 10 -6 mm/mm deg C, and the wet density between 2,610 and 2,820 kg/m 3 . The granite pegmatite has slightly lower coefficient of thermal expansion and wet density than gneissic rocks. (orig.)

Intense passionate love attenuates cigarette cue-reactivity in nicotine-deprived smokers: an FMRI study.

Directory of Open Access Journals (Sweden)

Xiaomeng Xu

Full Text Available Self-expanding experiences like falling in love or engaging in novel, exciting and interesting activities activate the same brain reward mechanism (mesolimbic dopamine pathway that reinforces drug use and abuse, including tobacco smoking. This suggests the possibility that reward from smoking is substitutable by self-expansion (through competition with the same neural system, potentially aiding cessation efforts. Using a model of self-expansion in the context of romantic love, the present fMRI experiment examined whether, among nicotine-deprived smokers, relationship self-expansion is associated with deactivation of cigarette cue-reactivity regions. Results indicated that among participants who were experiencing moderate levels of craving, cigarette cue-reactivity regions (e.g., cuneus and posterior cingulate cortex showed significantly less activation during self-expansion conditions compared with control conditions. These results provide evidence that rewards from one domain (self-expansion can act as a substitute for reward from another domain (nicotine to attenuate cigarette cue reactivity.

Intense passionate love attenuates cigarette cue-reactivity in nicotine-deprived smokers: an FMRI study.

Science.gov (United States)

Xu, Xiaomeng; Wang, Jin; Aron, Arthur; Lei, Wei; Westmaas, J Lee; Weng, Xuchu

2012-01-01

Self-expanding experiences like falling in love or engaging in novel, exciting and interesting activities activate the same brain reward mechanism (mesolimbic dopamine pathway) that reinforces drug use and abuse, including tobacco smoking. This suggests the possibility that reward from smoking is substitutable by self-expansion (through competition with the same neural system), potentially aiding cessation efforts. Using a model of self-expansion in the context of romantic love, the present fMRI experiment examined whether, among nicotine-deprived smokers, relationship self-expansion is associated with deactivation of cigarette cue-reactivity regions. Results indicated that among participants who were experiencing moderate levels of craving, cigarette cue-reactivity regions (e.g., cuneus and posterior cingulate cortex) showed significantly less activation during self-expansion conditions compared with control conditions. These results provide evidence that rewards from one domain (self-expansion) can act as a substitute for reward from another domain (nicotine) to attenuate cigarette cue reactivity.

Estimation of reactor core calculation by HELIOS/MASTER at power generating condition through DeCART, whole-core transport code

International Nuclear Information System (INIS)

Kim, H. Y.; Joo, H. G.; Kim, K. S.; Kim, G. Y.; Jang, M. H.

2003-01-01

The reactivity and power distribution errors of the HELIOS/MASTER core calculation under power generating conditions are assessed using a whole core transport code DeCART. For this work, the cross section tablesets were generated for a medium sized PWR following the standard procedure and two group nodal core calculations were performed. The test cases include the HELIOS calculations for 2-D assemblies at constant thermal conditions, MASTER 3D assembly calculations at power generating conditions, and the core calculations at HZP, HFP, and an abnormal power conditions. In all these cases, the results of the DeCART code in which pinwise thermal feedback effects are incorporated are used as the reference. The core reactivity, assemblywise power distribution, axial power distribution, peaking factor, and thermal feedback effects are then compared. The comparison shows that the error of the HELIOS/MASTER system in the core reactivity, assembly wise power distribution, pin peaking factor are only 100∼300 pcm, 3%, and 2%, respectively. As far as the detailed pinwise power distribution is concerned, however, errors greater than 15% are observed

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.

Adsorbate reactivity and thermal mobility from simple modeling of high-resolution core-level spectra: application to O/Al(111)

International Nuclear Information System (INIS)

Schouborg, Jakob; Raarup, Merete K; Balling, Peter

2009-01-01

A high-resolution core-level spectroscopy investigation of the adsorption of oxygen on Al(111) at variable oxygen exposure demonstrates a low surface reactivity for an intensively cleaned surface. The threshold for oxide formation is as high as ∼200 L (langmuirs), at which point the coverage of the chemisorbed oxygen exceeds half a monolayer. A simple model is presented, using which it is possible to deduce the oxygen coverage from the core-level spectra and determine the initial sticking probability. For our data a value of 0.018 ± 0.004 is obtained. The changes in core-level spectra following low-temperature annealing of low-coverage O/Al(111) reflect the formation of gradually larger islands of oxygen atoms (Ostwald ripening). The island formation is consistent with a random-walk model from which the diffusion barrier can be deduced to be in the range of 0.80-0.90 eV.

Bypass Flow and Hot Spot Analysis for PMR200 Block-Core Design with Core Restraint Mechanism

International Nuclear Information System (INIS)

Lim, Hong Sik; Kim, Min Hwan

2009-01-01

The accurate prediction of local hot spot during normal operation is important to ensure core thermal margin in a very high temperature gas-cooled reactor because of production of its high temperature output. The active cooling of the reactor core determining local hot spot is strongly affected by core bypass flows through the inter-column gaps between graphite blocks and the cross gaps between two stacked fuel blocks. The bypass gap sizes vary during core life cycle by the thermal expansion at the elevated temperature and the shrinkage/swelling by fast neutron irradiation. This study is to investigate the impacts of the variation of bypass gaps during core life cycle as well as core restraint mechanism on the amount of bypass flow and thus maximum fuel temperature. The core thermo fluid analysis is performed using the GAMMA+ code for the PMR200 block-core design. For the analysis not only are some modeling features, developed for solid conduction and bypass flow, are implemented into the GAMMA+ code but also non-uniform bypass gap distribution taken from a tool calculating the thermal expansion and the shrinkage/swell of graphite during core life cycle under the design options with and without core restraint mechanism is used

Design comparisons of TRU burner cores with similar sodium void worth

International Nuclear Information System (INIS)

Sang Ji, Kim; Young Il, Kim; Young Jin, Kim; Nam Zin, Cho

2001-01-01

This study summarizes the neutronic performance and fuel cycle behavior of five geometrically-different transuranic (TRU) burner cores with similar low sodium void reactivity. The conceptual cores encompass core geometries for annular, two-region homogeneous, dual pin type, pan-shaped and H-shaped cores. They have been designed with the same assembly specifications and managed to have similar end-of-cycle sodium void reactivities and beginning-of-cycle peak power densities through the changes in the core size and configuration. The requirement of low sodium void reactivity is shown to lead each design concept to characteristic neutronics performance and fuel cycle behavior. The H-/pan-shaped cores allow the core compaction as well as higher rate of TRU burning. (author)

Interactions of solitary waves and compression/expansion waves in core-annular flows

Science.gov (United States)

Maiden, Michelle; Anderson, Dalton; El, Gennady; Franco, Nevil; Hoefer, Mark

2017-11-01

The nonlinear hydrodynamics of an initial step leads to the formation of rarefaction waves and dispersive shock waves in dispersive media. Another hallmark of these media is the soliton, a localized traveling wave whose speed is amplitude dependent. Although compression/expansion waves and solitons have been well-studied individually, there has been no mathematical description of their interaction. In this talk, the interaction of solitons and shock/rarefaction waves for interfacial waves in viscous, miscible core-annular flows are modeled mathematically and explored experimentally. If the interior fluid is continuously injected, a deformable conduit forms whose interfacial dynamics are well-described by a scalar, dispersive nonlinear partial differential equation. The main focus is on interactions of solitons with dispersive shock waves and rarefaction waves. Theory predicts that a soliton can either be transmitted through or trapped by the extended hydrodynamic state. The notion of reciprocity is introduced whereby a soliton interacts with a shock wave in a reciprocal or dual fashion as with the rarefaction. Soliton reciprocity, trapping, and transmission are observed experimentally and are found to agree with the modulation theory and numerical simulations. This work was partially supported by NSF CAREER DMS-1255422 (M.A.H.) and NSF GRFP (M.D.M.).

Core-in-shell sorbent for hot coal gas desulfurization

Science.gov (United States)

Wheelock, Thomas D.; Akiti, Jr., Tetteh T.

2004-02-10

A core-in-shell sorbent is described herein. The core is reactive to the compounds of interest, and is preferably calcium-based, such as limestone for hot gas desulfurization. The shell is a porous protective layer, preferably inert, which allows the reactive core to remove the desired compounds while maintaining the desired physical characteristics to withstand the conditions of use.

Study of the core compaction effects and its monitoring in sodium cooled fast reactors

International Nuclear Information System (INIS)

Zylbersztejn, F.

2012-01-01

Conclusions: • On calculation of reactivity impacts of core compaction/flowering: → Upper bound of the reactivity coefficients for each type of deformation; → Uniform compaction model: significant reactivity impact; Circular symmetric model: small reactivity impact. • On the visibility of these phenomena by the neutron detectors: → The direct monitoring of the core compaction by neutron detector in the BCC is not possible. (the identification that the reactivity perturbations observed are due to variation of the core geometry). Perspectives of solutions: → Improved core design: reducing the effects. → Physical improvements: Steel resistance to deformations (irradiation, flexion); Direct devices: core constraint (prevents deformations). → Additional calculations: Considering more localized deformations; Advanced monitoring with neutron noise (in progress)

Reactivity monitoring during reactor-reloading operations

International Nuclear Information System (INIS)

Baumann, N.P.; Ahlfeld, C.F.; Ridgely, G.C.

1983-01-01

At the Savannah River Plant (SRP) reloading operations during shutdown present special considerations in reactivity monitoring and control. Large reactivity changes may occur during reloading operations because of the heterogeneous nature of some core designs. This paper describes an improved monitoring system

Reactivity costs in MARIA reactor

International Nuclear Information System (INIS)

Marcinkowska, Zuzanna E.; Pytel, Krzysztof M.; Frydrysiak, Andrzej

2017-01-01

Highlights: • The methodology for calculating consumed fuel cost of excess reactivity is proposed. • Correlation between time integral of the core excess reactivity and released energy. • Reactivity price gives number of fuel elements required for given excess reactivity. - Abstract: For the reactor operation at high power level and carrying out experiments and irradiations the major cost of reactor operation is the expense of nuclear fuel. In this paper the methodology for calculating consumed fuel cost-relatedness of excess reactivity is proposed. Reactivity costs have been determined on the basis of operating data. A number of examples of calculating the reactivity costs for processes such as: strong absorbing material irradiation, molybdenium-99 production, beryllium matrix poisoning and increased moderator temperature illustrates proposed method.

Transport-diffusion comparisons for small core LMFBR disruptive accidents

International Nuclear Information System (INIS)

Tomlinson, E.T.

1977-11-01

A number of numerical experiments were performed to assess the validity of diffusion theory for calculating the reactivity state of various small core LMFBR disrupted geometries. The disrupted configurations correspond, in general, to various configurations predicted by SAS3A for transient undercooling (TUC) and transient overpower (TOP) accidents for homogeneous cores and to the ZPPR-7 configurations for heterogeneous core. In all TUC cases diffusion theory was shown to be inadequate for the calculation of reactivity changes during core disassembly

An evolutionary analysis of genome expansion and pathogenicity in Escherichia coli.

Science.gov (United States)

Bohlin, Jon; Brynildsrud, Ola B; Sekse, Camilla; Snipen, Lars

2014-10-09

There are several studies describing loss of genes through reductive evolution in microbes, but how selective forces are associated with genome expansion due to horizontal gene transfer (HGT) has not received similar attention. The aim of this study was therefore to examine how selective pressures influence genome expansion in 53 fully sequenced and assembled Escherichia coli strains. We also explored potential connections between genome expansion and the attainment of virulence factors. This was performed using estimations of several genomic parameters such as AT content, genomic drift (measured using relative entropy), genome size and estimated HGT size, which were subsequently compared to analogous parameters computed from the core genome consisting of 1729 genes common to the 53 E. coli strains. Moreover, we analyzed how selective pressures (quantified using relative entropy and dN/dS), acting on the E. coli core genome, influenced lineage and phylogroup formation. Hierarchical clustering of dS and dN estimations from the E. coli core genome resulted in phylogenetic trees with topologies in agreement with known E. coli taxonomy and phylogroups. High values of dS, compared to dN, indicate that the E. coli core genome has been subjected to substantial purifying selection over time; significantly more than the non-core part of the genome (pcoli genome size correlated with estimated HGT size (pcoli are largely attained through HGT. No associations were found between selective pressures operating on the E. coli core genome, as estimated using relative entropy, and genome size (p~0.98). On a larger time frame, genome expansion in E. coli, which is significantly associated with the acquisition of virulence factors, appears to be independent of selective forces operating on the core genome.

Adjusted neutron spectra of STEK cores for reactivity calculations

International Nuclear Information System (INIS)

Dekker, J.W.M.; Dragt, J.B.; Janssen, A.J.; Heijboer, R.J.; Klippel, H.Th.

1978-02-01

Neutron flux and adjoint flux spectra form a pre-requisite in the analysis of reactivity worth data measured in the STEK facility. First, a survey of all available information about these spectra is given. Next a special application of a general adjustment method is described. This method has been used to obtain adjusted STEK group flux and adjoint flux spectra, starting from calculated spectra. These theoretical spectra were adjusted to reactivity worths of natural boron (nat. B) and 235 U as well as a number of fission reaction rates. As a by-product in this adjustment calculation adjusted fission group cross sections of 235 U were obtained. The results, viz. group fluxes and adjoint fluxes and adjusted fission cross sections of 235 U are given. They have been used for the interpretation of fission product reactivity worth measurements made in STEK

A Metal Fuel Core Concept for 1000 MWt Advanced Burner Reactor

International Nuclear Information System (INIS)

Yang, W.S.; Kim, T.K.; Grandy, C.

2007-01-01

This paper describes the core design and performance characteristics of a metal fuel core concept for a 1000 MWt Advanced Burner Reactor. A ternary metal fuel form of U-TRU-Zr was assumed with weapons grade plutonium feed for the startup core and TRU recovered from LWR spent fuel for the recycled equilibrium core. A compact burner core was developed by trade-off between the burnup reactivity loss and TRU conversion ratio, with a fixed cycle length of one-year. In the startup core, the average TRU enrichment is 15.5%, the TRU conversion ratio is 0.81, and the burnup reactivity loss over a cycle is 3.6% Δk. The heavy metal and TRU inventories are 13.1 and 2.0 metric tons, respectively. The average discharge burnup is 93 MWd/kg, and the TRU consumption rate is 55.5 kg/year. For the recycled equilibrium core, the average TRU enrichment is 22.1 %, the TRU conversion ratio is 0.73, and the burnup reactivity loss is 2.2% Δk. The TRU inventory and consumption rate are 2.9 metric tons and 81.6 kg/year, respectively. The evaluated reactivity coefficients provide sufficient negative feedbacks. The control systems provide shutdown margins that are more than adequate. The integral reactivity parameters for quasi-static reactivity balance analysis indicate favorable passive safety features, although detailed safety analyses are required to verify passive safety behavior. (authors)

Analyses of Design Extended Condition Events for the Prototype Gen-IV SFR

International Nuclear Information System (INIS)

Choi, Chiwoong; Jeong, Taekyung; Lee, Kwilim; Jeong, Jaeho; Ha, Kwiseok

2015-01-01

In this study, the sensitivity tests are conducted. In the case of the UTOP event, a sensitivity test for the reactivity insertion amount and rate were conducted. This analysis can give a requirement for margin of control rod stop system (CRSS). For example, the CRSS in the PRISM designed based on the 0.4 $ reactivity insertion, which is analyzed with safety analysis of UTOP event. Moreover, the sensitivity tests for weighting factor in the core radial expansion reactivity feedback model were also carried out for all ATWS events. Currently, the reactivity feedback model for the PGSFR is not validated yet. However, the reactivity feedback models in the MARS-LMR are validating with various plant-based data including EBR-II SHRT. The ATWS events for the PGSFR classified in the design extended condition including UTOP, ULOF, and ULOHS are analyzed with MARS-LMR. In this study, the sensitivity tests for reactivity insertion amount and rate in the UTOP event are conducted. The reactivity insertion amount is obviously an influential parameter. The reactivity insertion amount can give a requirement for design of the CRSS, therefore, this sensitivity result is very important to the CRSS. In addition, sensitivity tests for the weighting factor in the radial expansion reactivity model are carried out. The weighting factor for a grid plate, W GP , which means contribution of feedback in the grid plate is changed for all unprotected events. The grid plate expansion is governed by a core inlet temperature. As the W GP is increased, the power in the UTOP and the ULOF is increased, however, the power in the ULOHS is decreased. The higher power during transient means lower reactivity feedback and smaller expansion. Thus, the core outlet temperature rise is dominant in the UTOP and ULOF events, however, the core inlet temperature rise is dominant in the ULOHS. Therefore, the grid plate expansion in the ULOHS is predominant

Determination of the design excess reactivity for the TREAT Upgrade reactor

International Nuclear Information System (INIS)

Bhattacharyya, S.K.; Hanan, N.A.

1983-01-01

The excess reactivity designed to be built into a reactor core is a primary determinant of the fissile loadings of the fuel rods in the core. For the TREAT Upgrade (TU) reactor the considerations that enter into the determination of the excess reactivity are different from those of conventional power reactors. The reactor is designed to operate in an adiabatic transient mode for reactor safety in-pile test programs. The primary constituent of the excess reactivity is the calculated reactivity required to perform the most demanding transient experiments. Because of the unavailability of supporting critical experiments for the core design, the uncertainty terms that add on to this basic constituent are rather large. The burnup effects in TU are negligible and no refueling is planned. In this paper the determination of the design excess reactivity of the TREAT Upgrade reactor is discussed

Investigation of space-energy effects in the reactivity measurement by neutron noise with ex-core detectors in a reflected LWR

International Nuclear Information System (INIS)

Lescano, V.H.; Behringer, K.

1981-11-01

Practical application of the zero-crossing correlation method for measuring slightly subcritical reactivities in a swimming pool reactor required the use of detector locations in the reflector zone near to the core boundary. Experimental investigations of neutron-noise cross-power spectra showed significant deviations from the point reactor model at higher frequencies (> 100 Hz). Nevertheless, the use of the point reactor model was found to be an useful approach in the analysis of the zero-crossing correlation method yielding results which agreed well with those obtained from the rod-drop method. The theoretical part of the work is concerned with a space-dependent model calculation in two-group diffusion theory to support the experimental findings. The model calculation can explain the trends observed in the neutron-noise spectra as well as the applicability of the point reactor model to the zero-crossing correlation method. To obtain better insight, the calculations have been extended to neutron-noise spectra when one or both detectors are located in the core zone. In the case of a large core and widely spaced detectors, with at least one detector in the core zone, a sink frequency appears in the spectra. This effect is well-known in coupled-core kinetics. (Auth.)

Assessment of CANDU-6 reactivity devices for DUPIC fuel

International Nuclear Information System (INIS)

Jeong, Chang Joon; Choi, Hang Bok

1998-11-01

Reactivity device characteristics for a CANDU 6 reactor loaded with DUPIC fuel have been assessed. The lattice parameters were generated by WIMS-AECL code and the core calculations were performed by RFSP code with a 3-dimensional full core model. The reactivity devices studied are the zone controller, adjusters, mechanical control absorber and shutoff rods. For the zone controller system, damping capability for spatial oscillation was investigated. For the adjusters, the restart capability was investigated. For the adjusters, the restart capability was investigated. The shin operation and power stepback calculation were also performed to confirm the compatibility of the current adjuster system. The mechanical control absorber was assessed for the function of compensating temperature reactivity feedback following a power reduction. And shutoff rods were also assessed to investigate the following a power reduction. And shutoff rods were also assessed to investigate the static reactivity worth. This study has shown that the current reactivity device system of CANDU-6 core with the DUPIC fuel. (author). 9 refs., 17 tabs., 7 figs

Sensitivity of BWR shutdown margin tests to local reactivity anomalies

International Nuclear Information System (INIS)

Cokinos, D.M.; Carew, J.F.

1987-01-01

Successful shutdown margin (SDM) demonstration is a required procedure in the startup of a newly configured boiling water reactor (BWR) core. In its most reactive condition throughout a cycle, a BWR core must be capable of being made subcritical by a specified margin with the highest worth control rod fully withdrawn and all other rods at their fully inserted positions. Two different methods are used to demonstrate SDM: (a) the adjacent-rod test and (b) the in-sequence test. In the adjacent-rod test, the strongest rod is fully withdrawn and an adjacent rod is withdrawn to reach criticality. In the in-sequence test, control rods spread throughout the core are withdrawn in a predetermined sequence of withdrawals. Larger than expected core k/sub eff/ values have been observed during the performance of BWR SDM tests. The purpose of the work summarized in this paper has been to investigated and quantify the sensitivity of both the adjacent-rod and in-sequence SDM tests to local reactivity anomalies. This was accomplished by introducing reactivity perturbations at selected four-bundle cell locations and by evaluating their effect on core reactivity in each of the two tests

Feasibility study of SMART core with soluble boron

International Nuclear Information System (INIS)

Kim, Kang Seog; Lee, Chung Chan; Zee, Sung Quun

2000-11-01

The excess reactivity of SMART core without soluble boron is effectively controlled by 49 CEDM. We suggest another method to control the core excess reactivity using both the checkerboard type of 25 CEDM and soluble boron and perform a feasibility calculation. The soluble boron operation is categorized into the on-line and the off-line mechanisms. The former is to successively control the boron concentration according to the excess reactivity during operation and the latter is to add and change some soluble boron during refueling and repairing. Since the on-line soluble boron control system of SMART is conceptually identical to that of the commercial pressurized water reactor, we did not perform the analysis. Since the soluble boron in the complete off-line system increases the moderator temperature coefficient, the reactivity defect between hot and cold moderator temperature is decreased. However, the decrease of the reactivity is not big to satisfy the core reactivity limits. When using 25 CEDM, the possible mechanism is to control the excess reactivity by both control rod and on-line boron control mechanism between cold and hot zero power and by only control rod at hot full power. We selected the loading pattern satisfying the requirement in the view of nuclear design

Method of allowing for resonances in calculating reactivity values

International Nuclear Information System (INIS)

Kumpf, H.

1985-01-01

On the basis of the integral transport equation for the source density an expression has been derived for calculating reactivity values taking resonances in the core and in the sample into account. The model has been used for evaluating reactivities measured in the Rossendorf SEG IV configuration. It is shown that the influence of resonances in the core can be kept tolerable, if a sufficiently thick buffer zone of only slightly absorbing non-resonant material is arranged between the sample and the core. (author)

BN600 reactivity definition

International Nuclear Information System (INIS)

Zheltyshev, V.; Ivanov, A.

2000-01-01

Since 1980, the fast BN600 reactor with sodium coolant has been operated at Beloyarsk Nuclear Power Plant. The periodic monitoring of the reactivity modifications should be implemented in compliance with the standards and regulations applied in nuclear power engineering. The reactivity measurements are carried out in order to confirm the basic neutronic features of a BN600 reactor. The reactivity measurements are aimed to justify that nuclear safety is provided in course of the in-reactor installation of the experimental core components. Two reactivity meters are to be used on BN600 operation: 1. Digital on-line reactivity calculated under stationary reactor operation on power (approximation of the point-wise kinetics is applied). 2. Second reactivity meter used to define the reactor control rod operating components efficiency under reactor startup and take account of the changing efficiency of the sensor, however, this is more time-consumptive than the on-line reactivity meter. The application of two reactivity meters allows for the monitoring of the reactor reactivity under every operating mode. (authors)

Structure, Reactivity and Dynamics

Indian Academy of Sciences (India)

Understanding structure, reactivity and dynamics is the core issue in chemical ... functional theory (DFT) calculations, molecular dynamics (MD) simulations, light- ... between water and protein oxygen atoms, the superionic conductors which ...

CP ESFR: Collaborative Project for a European Sodium Fast Reactor Core studies

International Nuclear Information System (INIS)

Buiron, L.; Vasile, A.; Sunderland, R.

2013-01-01

• Significant progress has been made in optimizing both the oxide and carbide ESFR cores; • For the oxide core the optimisation process concentrated on the reduction of the sodium void reactivity effect and on the evaluation of MA burning performances. The CONF2 axial configuration has provided a significant overall reduction of the sodium void reactivity effect. • The carbide core had a significantly higher reactivity loss over the fuel cycle compared to the oxide one. By increasing slightly the fuel pin diameter, whilst still retaining the advantages of lower fuel temperatures of carbide fuel, and making changes in the core layout, the reactivity loss over the cycle has been reduced to a level similar to that of the oxide core. By adopting the CONF2 axial configuration initially developed for the oxide core, the sodium void reactivity of the carbide core has also been reduced appreciably. • The MA transmutation performances of the optimized ESFR oxide core have been investigated with respect to two boundary configurations. The HET2 configuration shows a low MA transmutation rate sufficient to burn the MA produced by the ESFR core without affecting the safety parameters. The HOM4 configuration (where 4%wt. MA are loaded homogeneously in each core SA) is the most challenging configuration due to its impact on safety coefficients but it shows an high MA burning rate suitable for burning also MA accumulated by a thermal reactor fleet

JOYO MK-II core characteristics database

International Nuclear Information System (INIS)

Tabuchi, Shiro; Aoyama, Takafumi; Nagasaki, Hideaki; Kato, Yuichi

1998-12-01

The experimental fast reactor JOYO served as the MK-II irradiation bed core for testing fuel and material for FBR development for 15 years from 1982 to 1997. During the MK-II operation, extensive data were accumulated from the core characteristics tests conducted in thirty-one duty operations and thirteen special test operations. These core management data and core characteristics data were compiled into a database. The code system MAGI has been developed and used for core management of JOYO MK-II, and the core characteristics and the irradiation test conditions were calculated using MAGI on the basis of three dimensional diffusion theory with seven neutron energy groups. The core management data include extensive data, which were recorded on CD-ROM for user convenience. The data are specifications and configurations of the core, and for about 300 driver fuel subassemblies and about 60 uninstrumented irradiation subassemblies are core composition before and after irradiation, neutron flux, neutron fluences, fuel and control rod burn-up, and temperature and power distributions. MK-II core characteristics and test conditions were stored in the database for post analysis. Core characteristics data include excess reactivities, control rod worths, and reactivity coefficients, e.g., temperature, power and burn-up. Test conditions include both measured and calculated data for irradiation conditions. (author)

An alternative approach to the management of reactive metals: tolerant cementitious systems

International Nuclear Information System (INIS)

Swift, P.; Cox, J.; Wise, M.; McKinney, J.; Rhodes, C.

2015-01-01

In recent years research has focused on preventing or minimising corrosion of reactive metals to ensure long-term waste package integrity. An alternative approach to the encapsulation of reactive metals is being explored. The approach will identify a cementitious-based encapsulating material that will allow corrosion of reactive metals to occur in a controlled and predictable manner, rather than seeking to limit or prevent the corrosion, whilst retaining waste package integrity. A low strength grout will be developed that will be 'tolerant' to the expansive forces generated by the corrosion products of reactive metals. Novel cementitious systems (e.g. foamed cements, rubber composite cements, cenosphere composite cements, lime mortars, bentonite cements etc.) that may be tolerant to potentially expansive waste products, such as reactive metals will be considered and assessed in a series of small-scale preliminary trials (compressive strength, porosity, permeability, pore solution pH, etc.)

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

Feasibility Study of Core Design with a Monte Carlo Code for APR1400 Initial core

Energy Technology Data Exchange (ETDEWEB)

Kim, Jinsun; Chang, Do Ik; Seong, Kibong [KEPCO NF, Daejeon (Korea, Republic of)

2014-10-15

The Monte Carlo calculation becomes more popular and useful nowadays due to the rapid progress in computing power and parallel calculation techniques. There have been many attempts to analyze a commercial core by Monte Carlo transport code using the enhanced computer capability, recently. In this paper, Monte Carlo calculation of APR1400 initial core has been performed and the results are compared with the calculation results of conventional deterministic code to find out the feasibility of core design using Monte Carlo code. SERPENT, a 3D continuous-energy Monte Carlo reactor physics burnup calculation code is used for this purpose and the KARMA-ASTRA code system, which is used for a deterministic code of comparison. The preliminary investigation for the feasibility of commercial core design with Monte Carlo code was performed in this study. Simplified core geometry modeling was performed for the reactor core surroundings and reactor coolant model is based on two region model. The reactivity difference at HZP ARO condition between Monte Carlo code and the deterministic code is consistent with each other and the reactivity difference during the depletion could be reduced by adopting the realistic moderator temperature. The reactivity difference calculated at HFP, BOC, ARO equilibrium condition was 180 ±9 pcm, with axial moderator temperature of a deterministic code. The computing time will be a significant burden at this time for the application of Monte Carlo code to the commercial core design even with the application of parallel computing because numerous core simulations are required for actual loading pattern search. One of the remedy will be a combination of Monte Carlo code and the deterministic code to generate the physics data. The comparison of physics parameters with sophisticated moderator temperature modeling and depletion will be performed for a further study.

Reactivity feedback coefficients Pakistan research reactor-1 using PRIDE code

Energy Technology Data Exchange (ETDEWEB)

Mansoor, Ali; Ahmed, Siraj-ul-Islam; Khan, Rustam [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Inam-ul-Haq [Comsats Institute of Information Technology, Islamabad (Pakistan). Dept. of Physics

2017-05-15

Results of the analyses performed for fuel, moderator and void's temperature feedback reactivity coefficients for the first high power core configuration of Pakistan Research Reactor - 1 (PARR-1) are summarized. For this purpose, a validated three dimensional model of PARR-1 core was developed and confirmed against the reference results for reactivity calculations. The ''Program for Reactor In-Core Analysis using Diffusion Equation'' (PRIDE) code was used for development of global (3-dimensional) model in conjunction with WIMSD4 for lattice cell modeling. Values for isothermal fuel, moderator and void's temperature feedback reactivity coefficients have been calculated. Additionally, flux profiles for the five energy groups were also generated.

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)

Design factors affecting dynamic behaviour of fast reactor cores. UK review paper

Energy Technology Data Exchange (ETDEWEB)

Brindley, K W [National Nuclear Corporation Ltd., Risley, Warrington (United Kingdom); Perks, M A [United Kingdom Atomic Energy Authority, Risley, Warrington (United Kingdom)

1982-01-01

This paper summarises the consideration that has been given in the UK to the following factors that affect the dynamic behaviour of fast reactor cores: fuel design - Pu/u homogeneity, fuel expansion, fuel-clad gaps, uranium fraction. Structural response - CR supports, diagrid, sub-assembly bowing sodium expansion coefficients - low void cores including heterogenous cores. Calculational methods and models are outlined and some experimental results are discussed. (author)

Core optimization studies for a small heating reactor

International Nuclear Information System (INIS)

Galperin, A.

1986-11-01

Small heating reactor cores are characterized by a high contribution of the leakage to the neutron balance and by a large power density variation in the axial direction. A limited number of positions is available for the control rods, which are necessary to satisfy overall reactivity requirements subject to a safety related constraint on the maximum worth of each rod. Design approaches aimed to improve safety and fuel utilization performance of the core include separation of the cooling and moderating functions of the water with the core in order to reduce hot-to-cold reactivity shift and judicious application of the axial Gd zoning aimed to improve the discharge burnup distribution. Several design options are analyzed indicating a satisfactory solution of the axial burnup distribution problem. The feasibility of the control rod system including zircaloy, stainless steel, natural boron and possibly enriched boron rods is demonstrated. A preliminary analysis indicates directions for further improvements of the core performance by an additional reduction of the hot-to-cold reactivity shift and by a reduction of the depletion reactivity swing adopting a higher gadolinium concentration in the fuel or a two-batch fuel management scheme. (author)

3D core burnup studies in 500 MWe Indian prototype fast breeder reactor to attain enhanced core burnup

International Nuclear Information System (INIS)

Choudhry, Nakul; Riyas, A.; Devan, K.; Mohanakrishnan, P.

2013-01-01

Highlights: ► Enhanced burnup potential of existing prototype fast breeder reactor core is studied. ► By increasing the Pu enrichment, fuel burnup can be increased in existing PFBR core. ► Enhanced burnup increase economy and reduce load of fuel fabrication and reprocessing. ► Beginning of life reactivity is suppressed by increasing the number of diluents. ► Absorber rod worth requirements can be achieved by increasing 10 B enrichment. -- Abstract: Fast breeder reactors are capable of producing high fuel burnup because of higher internal breeding of fissile material and lesser parasitic capture of neutrons in the core. As these reactors need high fissile enrichment, high fuel burnup is desirable to be cost effective and to reduce the load on fuel reprocessing and fabrication plants. A pool type, liquid sodium cooled, mixed (Pu–U) oxide fueled 500 MWe prototype fast breeder reactor (PFBR), under construction at Kalpakkam is designed for a peak burnup of 100 GWd/t. This limitation on burnup is purely due to metallurgical properties of structural materials like clad and hexcan to withstand high neutron fluence, and not by the limitation on the excess reactivity available in the core. The 3D core burnup studies performed earlier for approach to equilibrium core of PFBR is continued to demonstrate the burnup potential of existing PFBR core. To increase the fuel burnup of PFBR, plutonium oxide enrichment is increased from 20.7%/27.7% to 22.1%/29.4% of core-1/core-2 which resulted in cycle length increase from 180 to 250 effective full power days (efpd), so that the peak fuel burnup increases from 100 to 134 GWd/t, keeping all the core parameters under allowed safety limits. Number of diluents subassemblies is increased from eight to twelve at beginning of life core to bring down the initial core excess reactivity. PFBR refueling is revised to accommodate twelve diluents. Increase of 10 B enrichment in control safety rods (CSRs) and diverse safety rods (DSRs

Characteristic test of initial HTTR core

International Nuclear Information System (INIS)

Nojiri, Naoki; Shimakawa, Satoshi; Fujimoto, Nozomu; Goto, Minoru

2004-01-01

This paper describes the results of core physics test in start-up and power-up of the HTTR. The tests were conducted in order to ensure performance and safety of the high temperature gas cooled reactor, and was carried out to measure the critical approach, the excess reactivity, the shutdown margin, the control rod worth, the reactivity coefficient, the neutron flux distribution and the power distribution. The expected core performance and the required reactor safety characteristics were verified from the results of measurements and calculations

Reactivity margins in heavy water moderated production reactors

International Nuclear Information System (INIS)

Benton, F.D.

1981-11-01

The design of the reactor core and components of the heavy water moderated reactors at the Savannah River Plant (SFP) can be varied to produce a number of isotopes. For the past decade, the predominant reactor core design has been the enriched-depleted lattice. In this lattice, fuel assemblies of highly enriched uranium and target assemblies of depleted uranium, which produce plutonium, occupy alternate lattice positions. This heterogeneous lattice arrangement and a nonuniform control rod distribution result in a reactor core that requires sophisticated calculational methods for accurate reactivity margin and power distribution predictions. For maximum accuracy, techniques must exist to provide a base of observed data for the calculations. Frequent enriched-depleted lattice design changes are required as product demands vary. These changes provided incentive for the development of techniques to combine the results of calculations and observed reactivity data to accurately and conveniently monitor reactivity margins during operation

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

Reactivity and Power Distribution Management in LEU-loaded Linear B and BR

International Nuclear Information System (INIS)

Hartanto, Donny; Kim, Yonghee

2013-01-01

In this paper, the relatively high excess reactivity issue during the initial transitional period was addressed. The design target is to achieve a maximum excess reactivity of about 1.0 dollar to prevent the possibility of the prompt jump critical accident. The initial core is divided into 2 radial Zr-zones in order to reduce the excess reactivity. By doing this, the power profile at the BOC can also be flattened. After the optimum initial core configuration has been found, the blanket region is also divided into 2 radial Zr-zones in order to flatten the power distribution at EOC. The neutronic analyses were all performed using the Monte Carlo code McCARD with ENDF-B/VII.0 library. It was found that by using the concave Zr-zoning in the initial core of B and BR, the maximum excess reactivity can be effectively lowered. The radial power profile can also be successfully flattened by using the Zr-zoning and concave initial core. The concave concept deserves more investigations for better performances of the B and BR core

Detection of antibodies to hepatitis B core antigen using the Abbott ARCHITECT anti-HBc assay: analysis of borderline reactive sera.

Science.gov (United States)

Ollier, Laurence; Laffont, Catherine; Kechkekian, Aurore; Doglio, Alain; Giordanengo, Valérie

2008-12-01

Routine use of the automated chemiluminescent microparticle immunoassay Abbott ARCHITECT anti-HBc for diagnosis of hepatitis B is limited in case of borderline reactive sera with low signal close to the cut-off index. In order to determine the significance of anti-HBc detection when borderline reactivity occurs using the ARCHITECT anti-HBc assay, a comparative study was designed. 3540 serum samples collected over a 2-month period in the hospital of Nice were examined for markers of HBV infection (HBsAg, anti-HBs and anti-HBc). One hundred seven samples with sufficient volume and with borderline reactivity by the ARCHITECT assay were tested by two other anti-HBc assays, a microparticle enzyme immunoassay (MEIA, AxSYM Core, Abbott Laboratories, IL, USA) and an enzyme linked fluorescent assay (ELFA, VIDAS Anti-HBc Total II, bioMérieux, Lyon, France). Only 46 samples were confirmed by the AxSYM and the VIDAS assays. Additional serological information linked to patient history showed that the remaining samples (61) were false positives (11), had low titer of anti-HBc antibodies (13), or were inconclusive (37). This comparative study highlighted the existence of a grey zone around the cut-off index. Confirmative results through a different immunoassay are needed to confirm the diagnosis of HBV on borderline reactive sera using the ARCHITECT anti-HBc assay.

Relations among several nuclear and electronic density functional reactivity indexes

Science.gov (United States)

Torrent-Sucarrat, Miquel; Luis, Josep M.; Duran, Miquel; Toro-Labbé, Alejandro; Solà, Miquel

2003-11-01

An expansion of the energy functional in terms of the total number of electrons and the normal coordinates within the canonical ensemble is presented. A comparison of this expansion with the expansion of the energy in terms of the total number of electrons and the external potential leads to new relations among common density functional reactivity descriptors. The formulas obtained provide explicit links between important quantities related to the chemical reactivity of a system. In particular, the relation between the nuclear and the electronic Fukui functions is recovered. The connection between the derivatives of the electronic energy and the nuclear repulsion energy with respect to the external potential offers a proof for the "Quantum Chemical le Chatelier Principle." Finally, the nuclear linear response function is defined and the relation of this function with the electronic linear response function is given.

BN-600 hybrid core benchmark analyses

International Nuclear Information System (INIS)

Kim, Y.I.; Stanculescu, A.; Finck, P.; Hill, R.N.; Grimm, K.N.

2003-01-01

Benchmark analyses for the hybrid BN-600 reactor that contains three uranium enrichment zones and one plutonium zone in the core, have been performed within the frame of an IAEA sponsored Coordinated Research Project. The results for several relevant reactivity parameters obtained by the participants with their own state-of-the-art basic data and codes, were compared in terms of calculational uncertainty, and their effects on the ULOF transient behavior of the hybrid BN-600 core were evaluated. The comparison of the diffusion and transport results obtained for the homogeneous representation generally shows good agreement for most parameters between the RZ and HEX-Z models. The burnup effect and the heterogeneity effect on most reactivity parameters also show good agreement for the HEX-Z diffusion and transport theory results. A large difference noticed for the sodium and steel density coefficients is mainly due to differences in the spatial coefficient predictions for non fuelled regions. The burnup reactivity loss was evaluated to be 0.025 (4.3 $) within ∼ 5.0% standard deviation. The heterogeneity effect on most reactivity coefficients was estimated to be small. The heterogeneity treatment reduced the control rod worth by 2.3%. The heterogeneity effect on the k-eff and control rod worth appeared to differ strongly depending on the heterogeneity treatment method. A substantial spread noticed for several reactivity coefficients did not give a significant impact on the transient behavior prediction. This result is attributable to compensating effects between several reactivity effects and the specific design of the partially MOX fuelled hybrid core. (author)

Low reactivity penalty burnable poison rods

International Nuclear Information System (INIS)

1978-01-01

A nuclear reactor burnable poison rod is described which consists of an elongated tubular sheath enclosing a neutron absorbing material which, at least during reactor operation, also encloses a neutron moderating material. The excess reactivity existing at the beginning of core life is compensated for by the depletion of the burnable poison throughout the life of the core, so that the life of the core is extended. (UK)

Reactivity anomaly surveillance in the Fast Flux Test Facility through cycle 3

International Nuclear Information System (INIS)

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

1984-08-01

The technique for monitoring core reactivity during power operation used at the Fast Flux Test Facility (FFTF) is described. This technique relies on comparing predicted to measured rod positions to detect any anomalous (or unpredicted) core reactivity changes. It is implemented on the Plant Data System (PDS) computer and thus provides rapid indication of any abnormal core conditions. The prediction algorithms use thermal-hydraulic, control rod position and neutron flux sensor information to predict the core reactivity state. Initial results of using this technique based mainly on theoretical formulations is presented. The results show that the reactivity changes due to increasing reactor power (power defect) and burnup of the fuel were within approx. 16% of predicted values. To increase the sensitivity and accuracy of this technique, the prediction algorithms were calibrated to actual operating data. The work of calibrating this technique and the results of using the calibrated technique up through the third full operating cycle are summarized

Coupling of 3-D core computational codes and a reactor simulation software for the computation of PWR reactivity accidents induced by thermal-hydraulic transients

International Nuclear Information System (INIS)

Raymond, P.; Caruge, D.; Paik, H.J.

1994-01-01

The French CEA has recently developed a set of new computer codes for reactor physics computations called the Saphir system which includes CRONOS-2, a three-dimensional neutronic code, FLICA-4, a three-dimensional core thermal hydraulic code, and FLICA-S, a primary loops thermal-hydraulic transient computation code, which are coupled and applied to analyze a severe reactivity accident induced by a thermal hydraulic transient: the Steamline Break accident for a pressurized water reactor until soluble boron begins to accumulate in the core. The coupling of these codes has proved to be numerically stable. 15 figs., 7 refs

Nonlinear thermo-optical properties of two-layered spherical system of gold nanoparticle core and water vapor shell during initial stage of shell expansion

Directory of Open Access Journals (Sweden)

Astafyeva Liudmila

2011-01-01

Full Text Available Abstract Nonlinear thermo-optical properties of two-layered spherical system of gold nanoparticle core and water vapor shell, created under laser heating of nanoparticle in water, were theoretically investigated. Vapor shell expansion leads to decreasing up to one to two orders of magnitude in comparison with initial values of scattering and extinction of the radiation with wavelengths 532 and 633 nm by system while shell radius is increased up to value of about two radii of nanoparticle. Subsequent increasing of shell radius more than two radii of nanoparticle leads to rise of scattering and extinction properties of system over initial values. The significant decrease of radiation scattering and extinction by system of nanoparticle-vapor shell can be used for experimental detection of the energy threshold of vapor shell formation and investigation of the first stages of its expansion. PACS: 42.62.BE. 78.67. BF

Application of the neutron noise technique for measurement of reactivity for subcritical reactor RA-4

International Nuclear Information System (INIS)

Orso, J; Marenzana, A

2012-01-01

Reactor core RA-4 is divided into two parts that come together to start reactor. The reactor with core separate has the largest subcritical condition, this condition is more secure and therefore the reactor shutdown. In this paper measurements are made of the decay constant of the neutron prompt ' P ', using the α-Rossi and α-Feynman methods to calculate the reactivity of the reactor core for different positions. Both techniques are compared and reactivity is obtained for several position of the reactor core using the α-Rossi technical which is obtained a function that gives the reactivity depending on the separation of the core length. Both techniques are verified using a no multiplicative system. Reactivity values for different position of the core obtained by α-Rossi technique are: $[0 cm] = (-11+/-1) dollar, $[3 cm] = (-7+/-1) dollar, $[3.5 cm] (-5.5+/-0.8) dollar, $[4.2 cm] = (-3.8+/-0.3) dollar y $[4.5] = (-3.0+/-0.1) dollar (author)

Progress and problems in modelling HTR core dynamics

International Nuclear Information System (INIS)

Scherer, W.; Gerwin, H.

1991-01-01

In recent years greater effort has been made to establish theoretical models for HTR core dynamics. At KFA Juelich the TINTE (TIme dependent Neutronics and TEmperatures) code system has been developed, which is able to model the primary circuit of an HTR plant using modern numerical techniques and taking into account the mutual interference of the relevant physical variables. The HTR core is treated in 2-D R-Z geometry for both nucleonics and thermo-fluid-dynamics. 2-energy-group diffusion theory is used in the nuclear part including 6 groups of delayed neutron precursors and 14 groups of decay heat producers. Local and non-local heat sources are incorporated, thus simulating gamma ray transport. The thermo-fluid-dynamics module accounts for heterogeneity effects due to the pebble bed structure. Pipes and other components of the primary loop are modelled in 1-D geometry. Forced convection may be treated as well as natural convection in case of blower breakdown accidents. Validation of TINTE has started using the results of a comprehensive experimental program that has been performed at the Arbeitsgemeinschaft Versuchsreaktor GmbH (AVR) high temperature pebble bed reactor at Juelich. In the frame of this program power transients were initiated by varying the helium blower rotational speed or by moving the control rods. In most cases a good accordance between experiment and calculation was found. Problems in modelling the special AVR reactor geometry in two dimensions are described and suggestions for overcoming the uncertainties of experimentally determined control rod reactivities are given. The influence of different polynomial expansions of xenon cross sections to long term transients is discussed together with effects of burnup during that time. Up to now the TINTE code has proven its general applicability to operational core transients of HTR. The effects of water ingress on reactivity, fuel element corrosion and cooling gas properties are now being

Automatic optimization of core loading patterns to maximize cycle energy production within operational constraints

International Nuclear Information System (INIS)

Hobson, G.H.; Turinsky, P.J.

1986-01-01

Computational capability has been developed to automatically determine the core loading pattern which minimizes fuel cycle costs for a pressurized water reactor. Equating fuel cycle cost minimization with core reactivity maximization, the objective is to determine the loading pattern which maximizes core reactivity at end-of-cycle while satisfying the power peaking constraint throughout the cycle and region average discharge burnup limit. The method utilizes a two-dimensional, coarse mesh, finite difference scheme to evaluate core reactivity and fluxes for an initial reference loading pattern as a function of cycle burnup. First order perturbation theory is applied to determine the effects of assembly shuffling on reactivity, power distribution, and end-of-cycle burnup

Simulation of rod drop experiments in the initial cores of Loviisa and Mochovce

International Nuclear Information System (INIS)

Kaloinen, E.; Kyrki-Rajamaeki, R.; Wasastjerna, F.

1999-01-01

Interpretation of rod drop measurements during startup tests of the Loviisa reactors has earlier been studied with two-dimensional core calculations using a spatial prompt jump approximation. In these calculations the prediction for the reactivity meter reading was lower than the measured values by 25%. Another approach to solve the problem is simulation of the rod drop experiment with dynamic core calculations coupled with out of core calculations to estimate the response of ex-core ionization chambers for the reactivity meter. This report described the calculations performed with the three-dimensional dynamic code HEXTRAN for prediction of the reactivity meter readings in rod drop experiments in initial cores of the WWER-440 reactors. (Authors)

EBRPOCO - a program to calculate detailed contributions of power reactivity components of EBR-II

International Nuclear Information System (INIS)

Meneghetti, D.; Kucera, D.A.

1981-01-01

The EBRPOCO program has been developed to facilitate the calculations of the power coefficients of reactivity of EBR-II loadings. The program enables contributions of various components of the power coefficient to be delineated axially for every subassembly. The program computes the reactivity contributions of the power coefficients resulting from: density reduction of sodium coolant due to temperature; displacement of sodium coolant by thermal expansions of cladding, structural rods, subassembly cans, and lower and upper axial reflectors; density reductions of these steel components due to temperature; displacement of bond-sodium (if present) in gaps by differential thermal expansions of fuel and cladding; density reduction of bond-sodium (if present) in gaps due to temperature; free axial expansion of fuel if unrestricted by cladding or restricted axial expansion of fuel determined by axial expansion of cladding. Isotopic spatial contributions to the Doppler component my also be obtained. (orig.) [de

Analytical estimation of control rod shadowing effect for excess reactivity measurement of HTTR

International Nuclear Information System (INIS)

Nakano, Masaaki; Fujimoto, Nozomu; Yamashita, Kiyonobu

1999-01-01

The fuel addition method is generally used for the excess reactivity measurement of the initial core. The control rod shadowing effect for the excess reactivity measurement has been estimated analytically for High Temperature Engineering Test Reactor (HTTR). 3-dimensional whole core analyses were carried out. The movements of control rods in measurements were simulated in the calculation. It was made clear that the value of excess reactivity strongly depend on combinations of measuring control rods and compensating control rods. The differences in excess reactivity between combinations come from the control rod shadowing effect. The shadowing effect is reduced by the use of plural number of measuring and compensating control rods to prevent deep insertion of them into the core. The measured excess reactivity in the experiments is, however, smaller than the estimated value with shadowing effect. (author)

Development of small, fast reactor core designs using lead-based coolant

International Nuclear Information System (INIS)

Cahalan, J. E.; Hill, R. N.; Khalil, H. S.; Wade, D. C.

1999-01-01

A variety of small (100 MWe) fast reactor core designs are developed, these include compact configurations, long-lived (15-year fuel lifetime) cores, and derated, natural circulation designs. Trade studies are described which identify key core design issues for lead-based coolant systems. Performance parameters and reactivity feedback coefficients are compared for lead-bismuth eutectic (LBE) and sodium-cooled cores of consistent design. The results of these studies indicate that the superior neutron reflection capability of lead alloys reduces the enrichment and burnup swing compared to conventional sodium-cooled systems; however, the discharge fluence is significantly increased. The size requirement for long-lived systems is constrained by reactivity loss considerations, not fuel burnup or fluence limits. The derated lead-alloy cooled natural circulation cores require a core volume roughly eight times greater than conventional compact systems. In general, reactivity coefficients important for passive safety performance are less favorable for the larger, derated configurations

Evidence for coral range expansion accompanied by reduced diversity of Symbiodinium genotypes

KAUST Repository

Grupstra, Carsten G. B.

2017-05-15

Zooxanthellate corals are threatened by climate change but may be able to escape increasing temperatures by colonizing higher latitudes. To determine the effect of host range expansion on symbiont genetic diversity, we examined genetic variation among populations of Symbiodinium psygmophilum associated with Oculina patagonica, a range-expanding coral that acquires its symbionts through horizontal transmission. We optimized five microsatellite primer pairs for S. psygmophilum and tested them on Oculina spp. samples from the western North Atlantic and the Mediterranean. We then used them to compare symbiont genotype diversity between an Iberian core and an expansion front population of O. patagonica. Only one multilocus S. psygmophilum genotype was identified at the expansion front, and it was shared with the core population, which harbored seven multilocus genotypes. This pattern suggests that O. patagonica range expansion is accompanied by reduced symbiont genetic diversity, possibly due to limited dispersal of symbionts or local selection.

Measurements and calculation of reactivity in the IEA-R1 nuclear reactor

International Nuclear Information System (INIS)

Ferreira, P.S.B.

1988-01-01

Techniques and experimentals procedures utilized in the measurement of some nuclear parameters related to reactivity are presented. Measurements of reactivity coefficients, such as void, temperature and power, and control rod worth were made in the IEA-R1 Research Reactor. The techniques used to perform the measurements were: i) stable period (control rod calibration), ii) inverse kinetics (digital reactivity meter), iii) aluminium slab insertion in the fuel element coolant channels (void reactivity), iv) nuclear reactor core temperature changes by means of the changes in the coolant systems of reactor core (isothermal reactivity coefficient) and v) by making perturbation in the core through the control rod motions (power reactivity coefficient and control rod calibration). By using the computer codes HAMMER, HAMMER-TECHNION and CITATION, the experiments realized in the IEA-R1 reactor were simulated. From this simulation, the theoretical reactivity parameters were estimated and compared with the respective experimental results. Furthermore, in the second fuel load of Angra-1 Nuclear Power Station, the IPEN-CNEN/SP digital reactivity - meter were used in the lower power test with the aim to assess the equipment performance. Among several tests, the reacticity-meter were used in parallel with a Westinghouse analogic reativimeter-meter) to measure the heat additiona point, critical boron concentration, control rod calibration, isothermal and moderator reactivity coefficient. These tests, and the results obtained by the digital reactivity-meter are described. The results were compared with those obtained by Westinghouse analogic reactivity meter, showing excellent agreement. (author) [pt

IAEA sodium void reactivity benchmark calculations

International Nuclear Information System (INIS)

Hill, R.N.; Finck, P.J.

1992-01-01

In this paper, the IAEA-1 992 ''Benchmark Calculation of Sodium Void Reactivity Effect in Fast Reactor Core'' problem is evaluated. The proposed design is a large axially heterogeneous oxide-fueled fast reactor as described in Section 2; the core utilizes a sodium plenum above the core to enhance leakage effects. The calculation methods used in this benchmark evaluation are described in Section 3. In Section 4, the calculated core performance results for the benchmark reactor model are presented; and in Section 5, the influence of steel and interstitial sodium heterogeneity effects is estimated

BN-600 MOX Core Benchmark Analysis. Results from Phases 4 and 6 of a Coordinated Research Project on Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects

International Nuclear Information System (INIS)

2013-12-01

For those Member States that have or have had significant fast reactor development programmes, it is of utmost importance that they have validated up to date codes and methods for fast reactor physics analysis in support of R and D and core design activities in the area of actinide utilization and incineration. In particular, some Member States have recently focused on fast reactor systems for minor actinide transmutation and on cores optimized for consuming rather than breeding plutonium; the physics of the breeder reactor cycle having already been widely investigated. Plutonium burning systems may have an important role in managing plutonium stocks until the time when major programmes of self-sufficient fast breeder reactors are established. For assessing the safety of these systems, it is important to determine the prediction accuracy of transient simulations and their associated reactivity coefficients. In response to Member States' expressed interest, the IAEA sponsored a coordinated research project (CRP) on Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects. The CRP started in November 1999 and, at the first meeting, the members of the CRP endorsed a benchmark on the BN-600 hybrid core for consideration in its first studies. Benchmark analyses of the BN-600 hybrid core were performed during the first three phases of the CRP, investigating different nuclear data and levels of approximation in the calculation of safety related reactivity effects and their influence on uncertainties in transient analysis prediction. In an additional phase of the benchmark studies, experimental data were used for the verification and validation of nuclear data libraries and methods in support of the previous three phases. The results of phases 1, 2, 3 and 5 of the CRP are reported in IAEA-TECDOC-1623, BN-600 Hybrid Core Benchmark Analyses, Results from a Coordinated Research Project on Updated Codes and Methods to Reduce the

Establishing the long-term fuel management scheme using point reactivity model

International Nuclear Information System (INIS)

Park, Yong-Soo; Kim, Jae-Hak; Lee, Young-Ouk; Song, Jae-Woong; Zee, Sung-Kyun

1994-01-01

A new approach to establish the long-term fuel management scheme is presented in this paper. The point reactivity model is used to predict the core average reactivity. An attempt to calculate batchwise power fraction is introduced through the two-dimensional nodal power algorithm based on the modified one-group diffusion equation and the number of fuel assemblies on the core periphery. Suggested is an empirical formula to estimate the radial leakage reactivity with ripe core design experience reflected. This approach predicts the cycle lengths and the discharge burnups of individual fuel batches up to an equilibrium core when the proper input data such as batch enrichment, batch size, type and content of burnable poison and reloading strategies are given. Eight benchmark calculations demonstrate that the new approach used in this study is reasonably accurate and highly efficient for the purpose of scoping calculation when compared with design code predictions. (author)

Analysis of void reactivity measurements in full MOX BWR physics experiments

International Nuclear Information System (INIS)

Ando, Yoshihira; Yamamoto, Toru; Umano, Takuya

2008-01-01

In the full MOX BWR physics experiments, FUBILA, four 9x9 test assemblies simulating BWR full MOX assemblies were located in the center of the core. Changing the in-channel moderator condition of the four assemblies from 0% void to 40% and 70% void mock-up, void reactivity was measured using Amplified Source Method (ASM) technique in the subcritical cores, in which three fission chambers were located. ASM correction factors necessary to express the consistency of the detector efficiency between measured core configurations were calculated using collision probability cell calculation and 3D-transport core calculation with the nuclear data library, JENDL-3.3. Measured reactivity worth with ASM correction factor was compared with the calculated results obtained through a diffusion, transport and continuous energy Monte Carlo calculation respectively. It was confirmed that the measured void reactivity worth was reproduced well by calculations. (author)

Application of a Virtual Reactivity Feedback Control Loop in Non-Nuclear Testing of a Fast Spectrum Reactor

International Nuclear Information System (INIS)

Bragg-Sitton, Shannon M.; Forsbacka, Matthew

2004-01-01

For a compact, fast-spectrum reactor, reactivity feedback is dominated by core deformation at elevated temperature. Given the use of accurate deformation measurement techniques, it is possible to simulate nuclear feedback in non-nuclear electrically heated reactor tests. Implementation of simulated reactivity feedback in response to measured deflection is being tested at the Nasa Marshall Space Flight Center Early Flight Fission Test Facility (EFF-TF). During tests of the SAFE-100 reactor prototype, core deflection was monitored using a high resolution camera. 'Virtual' reactivity feedback was accomplished by applying the results of Monte Carlo calculations (MCNPX) to core deflection measurements; the computational analysis was used to establish the reactivity worth of various core deformations. The power delivered to the SAFE-100 prototype was then adjusted accordingly via kinetics calculations. The work presented in this paper will demonstrate virtual reactivity feedback as core power was increased from 1 kWt to 10 kWt, held approximately constant at 10 kWt, and then allowed to decrease based on the negative thermal reactivity coefficient. (authors)

An Interval Bound Algorithm of optimizing reactor core loading pattern by using reactivity interval schema

International Nuclear Information System (INIS)

Gong Zhaohu; Wang Kan; Yao Dong

2011-01-01

Highlights: → We present a new Loading Pattern Optimization method - Interval Bound Algorithm (IBA). → IBA directly uses the reactivity of fuel assemblies and burnable poison. → IBA can optimize fuel assembly orientation in a coupled way. → Numerical experiment shows that IBA outperforms genetic algorithm and engineers. → We devise DDWF technique to deal with multiple objectives and constraints. - Abstract: In order to optimize the core loading pattern in Nuclear Power Plants, the paper presents a new optimization method - Interval Bound Algorithm (IBA). Similar to the typical population based algorithms, e.g. genetic algorithm, IBA maintains a population of solutions and evolves them during the optimization process. IBA acquires the solution by statistical learning and sampling the control variable intervals of the population in each iteration. The control variables are the transforms of the reactivity of fuel assemblies or the worth of burnable poisons, which are the crucial heuristic information for loading pattern optimization problems. IBA can deal with the relationship between the dependent variables by defining the control variables. Based on the IBA algorithm, a parallel Loading Pattern Optimization code, named IBALPO, has been developed. To deal with multiple objectives and constraints, the Dynamic Discontinuous Weight Factors (DDWF) for the fitness function have been used in IBALPO. Finally, the code system has been used to solve a realistic reloading problem and a better pattern has been obtained compared with the ones searched by engineers and genetic algorithm, thus the performance of the code is proved.

JOYO MK-II core characteristics database

International Nuclear Information System (INIS)

Ohkawachi, Yasushi; Maeda, Shigetaka; Sekine, Takashi; Aoyama, Takafumi

2003-04-01

The 'JOYO' MK-II core characteristics database was compiled and published in 1998. Comments and requests from many users led to the creation of a revised edition. The revisions include changes to the MAGI calculation code system to use the 70 group JFS-3-J3.2 constant set processed from the JENDL-3.2 library. Total control rod worth, reactor kinetic parameters and the MK-II core performance test results were included per user's requests. The core characteristics obtained from the 32 nd to 35 th operational cycles, which were conducted in the MK-III transition core, were newly added in this revised version. The MK-II core management data and core characteristics data were recorded to CD-ROM for user convenience. The Configuration Data' include the core arrangement and refueling record for each operational cycle. The 'Subassembly Library Data' include the atomic number density, neutron fluence, burn-up, integral power of 362 driver fuel subassemblies and 69 irradiation test subassemblies. The 'Output Data' contain the calculated neutron flux, gamma flux, power density, linear heat rate, coolant and fuel temperature distribution of all the fuel subassemblies at the beginning and end of each operational cycle. The 'Core Characteristics Data' include the measured excess reactivity, control rod worth calibration curve, and reactivity coefficients of temperature, power and burn-up. (author)

INDIAN POINT REACTOR REACTIVITY AND FLUX DISTRIBUTION MEASUREMENTS

Energy Technology Data Exchange (ETDEWEB)

Batch, M. L.; Fischer, F. E.

1963-11-15

The reactivity of the Indian Point core was measured near zero reactivity at various shim and control rod patterns. Flux distribution measurements were also made, and the results are expressed in terms of power peaking factors and normalized detector response during rod withdrawal. (D.L.C.)

Thermal expansion of epoxy-fiberglass composite specimens

International Nuclear Information System (INIS)

McElroy, D.L.; Weaver, F.J.; Bridgman, C.

1986-01-01

The thermal expansion behavior of three epoxy-fiberglass composite specimens was measured from 20 to 120 0 C (70 to 250 0 F) using a fused quartz push-rod dilatometer. Billets produced by vacuum impregnating layers of two types of fiberglass cloth with an epoxy resin were core-drilled to produce cylindrical specimens. These were used to study expansion perpendicular and parallel to the fiberglass layers. The dilatometer is held at a preselected temperature until steady-state is indicated by stable length and temperature data. Before testing the composite specimens, a reliability check of the dilatometer was performed using a copper secondary standard. This indicated thermal expansion coefficient (α) values within +-2% of expected values from 20 to 200 0 C

The whole-core LEU silicide fuel demonstration in the JMTR

Energy Technology Data Exchange (ETDEWEB)

Aso, Tomokazu; Akashi, Kazutomo; Nagao, Yoshiharu [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)] [and others

1997-08-01

The JMTR was fully converted to LEU silicide (U{sub 3}Si{sub 2}) fuel with cadmium wires as burnable absorber in January, 1994. The reduced enrichment program for the JMTR was initiated in 1979, and the conversion to MEU (enrichment ; 45%) aluminide fuel was carried out in 1986 as the first step of the program. The final goal of the program was terminated by the present LEU conversion. This paper describes the results of core physics measurement through the conversion phase from MEU fuel core to LEU fuel core. Measured excess reactivities of the LEU fuel cores are mostly in good agreement with predicted values. Reactivity effect and burnup of cadmium wires, therefore, were proved to be well predicted. Control rod worth in the LEU fuel core is mostly less than that in the MEU fuel core. Shutdown margin was verified to be within the safety limit. There is no significant difference in temperature coefficient of reactivity between the MEU and LEU fuel cores. These results verified that the JMTR was successfully and safely converted to LEU fuel. Extension of the operating cycle period was achieved and reduction of spend fuel elements is expected by using the fuel with high uranium density.

THE IMPACT OF POWER COEFFICIENT OF REACTIVITY ON CANDU 6 REACTORS

Directory of Open Access Journals (Sweden)

D. KASTANYA

2013-10-01

Full Text Available The combined effects of reactivity coefficients, along with other core nuclear characteristics, determine reactor core behavior in normal operation and accident conditions. The Power Coefficient of Reactivity (PCR is an aggregate indicator representing the change in reactor core reactivity per unit change in reactor power. It is an integral quantity which captures the contributions of the fuel temperature, coolant void, and coolant temperature reactivity feedbacks. All nuclear reactor designs provide a balance between their inherent nuclear characteristics and the engineered reactivity control features, to ensure that changes in reactivity under all operating conditions are maintained within a safe range. The CANDU® reactor design takes advantage of its inherent nuclear characteristics, namely a small magnitude of reactivity coefficients, minimal excess reactivity, and very long prompt neutron lifetime, to mitigate the demand on the engineered systems for controlling reactivity and responding to accidents. In particular, CANDU reactors have always taken advantage of the small value of the PCR associated with their design characteristics, such that the overall design and safety characteristics of the reactor are not sensitive to the value of the PCR. For other reactor design concepts a PCR which is both large and negative is an important aspect in the design of their engineered systems for controlling reactivity. It will be demonstrated that during Loss of Regulation Control (LORC and Large Break Loss of Coolant Accident (LBLOCA events, the impact of variations in power coefficient, including a hypothesized larger than estimated PCR, has no safety-significance for CANDU reactor design. Since the CANDU 6 PCR is small, variations in the range of values for PCR on the performance or safety of the reactor are not significant.

Validation study of core analysis methods for full MOX BWR

International Nuclear Information System (INIS)

2013-01-01

JNES has been developing a technical database used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical database, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The database will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects: (1) improving a Doppler reactivity analysis model in a Monte Carlo calculation code MVP, (2) sensitivity study of nuclear cross section date on reactivity calculation of experimental cores composed of UO 2 and MOX fuel rods, (3) analysis of isotopic composition data for UO 2 and MOX fuels and (4) the guide of reviewing the core analysis codes and others. (author)

Validation study of core analysis methods for full MOX BWR

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

JNES has been developing a technical database used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical database, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The database will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects: (1) improving a Doppler reactivity analysis model in a Monte Carlo calculation code MVP, (2) sensitivity study of nuclear cross section date on reactivity calculation of experimental cores composed of UO{sub 2} and MOX fuel rods, (3) analysis of isotopic composition data for UO{sub 2} and MOX fuels and (4) the guide of reviewing the core analysis codes and others. (author)

Surface zwitterionicalization of poly(vinylidene fluoride) membranes from the entrapped reactive core-shell silica nanoparticles.

Science.gov (United States)

Zhu, Li-Jing; Zhu, Li-Ping; Zhang, Pei-Bin; Zhu, Bao-Ku; Xu, You-Yi

2016-04-15

We demonstrate the preparation and properties of poly(vinylidene fluoride) (PVDF) filtration membranes modified via surface zwitterionicalization mediated by reactive core-shell silica nanoparticles (SiO2 NPs). The organic/inorganic hybrid SiO2 NPs grafted with poly(methyl meth acrylate)-block-poly(2-dimethylaminoethyl methacrylate) copolymer (PMMA-b-PDMAEMA) shell were prepared by surface-initiated reversible addition fragmentation chain transfer (SI-RAFT) polymerization and then used as a membrane-making additive of PVDF membranes. The PDMAEMA exposed on membrane surface and pore walls were quaternized into zwitterionic poly(sulfobetaine methacrylate) (PSBMA) using 1,3-propane sultone (1,3-PS) as the quaternization agent. The membrane surface chemistry and morphology were analyzed by attenuated total reflectance Fourier transform infrared spectroscopy (ATR-FTIR), X-ray photoelectron spectroscopy (XPS) and scanning electron microscopy (SEM), respectively. The hydrophilicity, permeability and antifouling ability of the investigated membranes were evaluated in detail. It was found that the PSBMA chains brought highly-hydrophilic and strong fouling resistant characteristics to PVDF membranes due to the powerful hydration of zwitterionic surface. The SiO2 cores and PMMA chains in the hybrid NPs play a role of anchors for the linking of PSBMA chains to membrane surface. Compared to the traditional strategies for membrane hydrophilic modification, the developed method in this work combined the advantages of both blending and surface reaction. Copyright © 2016 Elsevier Inc. All rights reserved.

Thermal hydraulic And RSG-Gas Core Reactivity Characteristics Due To Cold Water Insertion Accident

International Nuclear Information System (INIS)

Hastuti, Endiah Puji; Suparlina, Lily; Tukiran

2000-01-01

Under normal operating condition,the primary coolant is circulated by 2 out of the 3 primary coolant pumps. Unnecessary operation of the reserve pump would result in a temperatur decrease of the primary coolant by less than 5 o C. the corresponding increase of reactivity amounts to Δρ ≤0,1 %. The analysis was done using silicide core configuration data with 3.55 gU /cm 3 fuel loading. The calculation model was done with and without automatic control rod. The calculation results for the worst case condition, shows that reactor reached the maximum power 28.52 MW at 81.1 seconds, after the accident occurred. The maximal fuel element, cladding and outlet coolant temperatures are 148.3 o C,142.1 o C, and 75.7 o C, respectively. Safety margins for DNBR and flow instability reached 1.25 and 4.20, respectively. Comparing to the RSG-GAS safety margin at transient condition reguirement >1.48, RSG-GAS has enough safety margin if the power trip executed at 114% of 25 MW

Detonative propagation and accelerative expansion of the Crab Nebula shock front.

Science.gov (United States)

Gao, Yang; Law, Chung K

2011-10-21

The accelerative expansion of the Crab Nebula's outer envelope is a mystery in dynamics, as a conventional expanding blast wave decelerates when bumping into the surrounding interstellar medium. Here we show that the strong relativistic pulsar wind bumping into its surrounding nebula induces energy-generating processes and initiates a detonation wave that propagates outward to form the current outer edge, namely, the shock front, of the nebula. The resulting detonation wave, with a reactive downstream, then provides the needed power to maintain propagation of the shock front. Furthermore, relaxation of the curvature-induced reduction of the propagation velocity from the initial state of formation to the asymptotic, planar state of Chapman-Jouguet propagation explains the observed accelerative expansion. Potential richness in incorporating reactive fronts in the description of various astronomical phenomena is expected. © 2011 American Physical Society

Light Water Breeder Reactor (LWBR) flow coefficient of reactivity: (LWBR Development Program)

International Nuclear Information System (INIS)

Sarber, W.K.; Stout, J.W.; Atherton, R.

1987-06-01

This report discusses the results of an experimental program to measure and categorize the causes for increases in the magnitude of the LWBR flow coefficient of reactivity at 10,932 EFPH from previously measured near zero values to a value of about 6 x 10 -4 Δ rho for a flow decrease from 100 to 80% of full flow. Reactor protection analyses confirmed that existing protection systems were adequate for continued operation. Subsequently, the flow coefficient decreased in magnitude to approximately 2.25 x 10 -4 Δ rho at 20,000 EFPH and remained about constant through the remainder of core life, 29,047 EFPH. The increase in flow coefficient of reactivity is attributed to a flow-force dependent change in the effective core diameter such that an increase in core flow decreased the core diameter, resulting in an increase in fuel-to-water ratio and a consequent decrease in the reactivity of this relatively undermoderated core. This report discusses why the increased flow coefficient did not occur until after 10,932 EFPH and why the magnitude of flow coefficient reduced with continued core operation

A reactivity hold-down strategy for soluble boron free operation by introducing Pu-238 added fuel

International Nuclear Information System (INIS)

Kim, Soon Young; Kim, Jong Kyung

2000-01-01

A new concept of Pu-238 added fuel is introduced to control the reactivity and power distribution in soluble boron free (SBF) pressurized water reactor (PWR) core. Though extensive use of burnable poison and control rods is inevitable for reactivity suppression in SBF core, it causes the core power distribution control to be so difficult that a practical SBF operation is far distant. In this work, it is confirmed that the excess reactivity can be greatly suppressed by introducing the Pu-238 added fuel. As a result of the conceptual core design of the 600 MWe SBF PWR using Pu-238 added fuel, the core reactivity is well controlled in comparison with the results obtained from the earlier 600 MWe SBF core design works. Especially, the axial power shape control is performed successfully with the aid of simple axial zoning scheme, developed in this study, by using Pu-238 enrichment zoning. The Pu-238 added fuel is also tested for 1300 MWe SBF PWR core design, in which the power distribution control can be more difficult than that of smaller plants if soluble boron control is not available. The results show that the core excess reactivity and the power distribution can be well controlled without using soluble boron even in a large-sized PWR. Hence, one of the difficult control problems arising in SBF core design can be greatly mitigated by introducing the new fuel concept. It is further expected that the Pu-238 added fuel, the simple axial zoning scheme, and the control bank operation strategy introduced in this study are directly applicable to practical SBF core design

Reactive diffusion and stresses in nanowires or nanorods

International Nuclear Information System (INIS)

Roussel, Manuel; Erdélyi, Zoltán; Schmitz, Guido

2017-01-01

Heterostructured nanowires are of prime interest in nowadays technology such as field-effect transistors, field emitters, batteries and solar cells. We consider their aging behavior and developed a model focusing on reactive diffusion in core-shell nanowires. A complete set of analytical equations is presented that takes into account thermodynamic driving forces, vacancy distribution, elastic stress and its plastic relaxation. This complete description of the reactive diffusion can be used in finite element simulations to investigate diffusion processes in various geometries. In order to show clearly the interplay between the cylindrical geometry, the reactive diffusion and the stresses developing in the nanowire, we investigate the formation of an intermetallic reaction product in various core-shell geometries. Emphasis is placed on showing how it is possible to control the kinetics of the reaction by applying an axial stress to the nanowires.

Aspects of cell calculations in deterministic reactor core analysis

International Nuclear Information System (INIS)

Varvayanni, M.; Savva, P.; Catsaros, N.

2011-01-01

Τhe capability of achieving optimum utilization of the deterministic neutronic codes is very important, since, although elaborate tools, they are still widely used for nuclear reactor core analyses, due to specific advantages that they present compared to Monte Carlo codes. The user of a deterministic neutronic code system has to make some significant physical assumptions if correct results are to be obtained. A decisive first step at which such assumptions are required is the one-dimensional cell calculations, which provide the neutronic properties of the homogenized core cells and collapse the cross sections into user-defined energy groups. One of the most crucial determinations required at the above stage and significantly influencing the subsequent three-dimensional calculations of reactivity, concerns the transverse leakages, associated to each one-dimensional, user-defined core cell. For the appropriate definition of the transverse leakages several parameters concerning the core configuration must be taken into account. Moreover, the suitability of the assumptions made for the transverse cell leakages, depends on earlier user decisions, such as those made for the core partition into homogeneous cells. In the present work, the sensitivity of the calculated core reactivity to the determined leakages of the individual cells constituting the core, is studied. Moreover, appropriate assumptions concerning the transverse leakages in the one-dimensional cell calculations are searched out. The study is performed examining also the influence of the core size and the reflector existence, while the effect of the decisions made for the core partition into homogenous cells is investigated. In addition, the effect of broadened moderator channels formed within the core (e.g. by removing fuel plates to create space for control rod hosting) is also examined. Since the study required a large number of conceptual core configurations, experimental data could not be available for

Characterization of Coated Sand Cores from Two Different Binder Systems for Grey Iron Castings

DEFF Research Database (Denmark)

Nwaogu, Ugochukwu Chibuzoh; Tiedje, Niels Skat; Poulsen, Thomas

or veining and metal penetration defects. The use of refractory coatings on cores is fundamental to obtaining acceptable casting surface quality and is used on resin bonded cores in production foundries. In this study new sol gel-coated sand cores made from coldbox and furan binder systems were investigated......Expansion defects on the surface of the castings include sand burn-in, metal penetration and/or veining, finning or scab. Veining or finning and metal penetration are of interest. These defects are associated with silica sand and result from the penetration of liquid metal into cracks formed during...... differential expansion of the core during heating. The rapid expansion of silica sand up to 600 oC and especially at 573 oC, where the α – β phase transformation occurs, is the cause of stresses in the core system. These stresses cause crack formation and metal melt flows into these cracks causing finning...

Arrival and expansion of the invasive foraminifera Trochammina hadai Uchio in Padilla Bay, Washington

Science.gov (United States)

McGann, Mary; Grossman, Eric E.; Takesue, Renee K.; Penttila, Dan; Walsh, John P.; Corbett, Reide

2012-01-01

Trochammina hadai Uchio, a benthic foraminifera native to Japanese estuaries, was first identified as an invasive in 1995 in San Francisco Bay and later in 16 other west coast estuaries. To investigate the timing of the arrival and expansion of this invasive species in Padilla Bay, Washington, we analyzed the distribution of foraminifera in two surface samples collected in 1971, in nine surface samples collected by Scott in 1972–1973, as well as in two cores (Padilla Flats 3 and Padilla V1/V2) obtained in 2004. Trochanimina hadai, originally identified as the native Trochammina pacifica Cushman in several early foraminiferal studies, dominates the assemblage of most of the surface samples. In the Padilla V1/V2 and Padilla Flats 3 cores, the species' abundance follows a pattern of absence, first appearance, rapid expansion commonly seen shortly after the arrival of a successful biological invasion, setback, and second expansion. Using Q-mode cluster analysis, pre-expansion and expansion assemblages were identified. Pb-210 dating of these cores proved unsuccessful. However, based on T. hadai's first appearance occurring stratigraphically well above sedimentological changes in the cores that reflect deposition of sediments in the bay due to previous diversions of the Skagit River, and its dominance in the early 1970s surface samples, we conclude that the species arrived in Padilla Bay somewhere between the late 1800s and 1971. Trochammina hadai may have been introduced into the bay in the 1930s when oyster culturing began there or, at a minimum, ten years prior to its appearance in San Francisco Bay.

Experimental Investigation of Reynolds Number Effects on Test Quality in a Hypersonic Expansion Tube

Science.gov (United States)

Rossmann, Tobias; Devin, Alyssa; Shi, Wen; Verhoog, Charles

2017-11-01

Reynolds number effects on test time and the temporal and spatial flow quality in a hypersonic expansion tube are explored using high-speed pressure, infrared optical, and Schlieren imaging measurements. Boundary layer models for shock tube flows are fairly well established to assist in the determination of test time and flow dimensions at typical high enthalpy test conditions. However, the application of these models needs to be more fully explored due to the unsteady expansion of turbulent boundary layers and contact regions separating dissimilar gasses present in expansion tube flows. Additionally, expansion tubes rely on the development of a steady jet with a large enough core-flow region at the exit of the acceleration tube to create a constant velocity region inside of the test section. High-speed measurements of pressure and Mach number at several locations within the expansion tube allow for the determination of an experimental x-t diagram. The comparison of the experimentally determined x-t diagram to theoretical highlights the Reynolds number dependent effects on expansion tube. Additionally, spatially resolved measurements of the Reynolds number dependent, steady core-flow in the expansion tube viewing section are shown. NSF MRI CBET #1531475, Lafayette College, McCutcheon Foundation.

In-core Instrument Subcritical Verification (INCISV) - Core Design Verification Method - 358

International Nuclear Information System (INIS)

Prible, M.C.; Heibel, M.D.; Conner, S.L.; Sebastiani, P.J.; Kistler, D.P.

2010-01-01

According to the standard on reload startup physics testing, ANSI/ANS 19.6.1, a plant must verify that the constructed core behaves sufficiently close to the designed core to confirm that the various safety analyses bound the actual behavior of the plant. A large portion of this verification must occur before the reactor operates at power. The INCISV Core Design Verification Method uses the unique characteristics of a Westinghouse Electric Company fixed in-core self powered detector design to perform core design verification after a core reload before power operation. A Vanadium self powered detector that spans the length of the active fuel region is capable of confirming the required core characteristics prior to power ascension; reactivity balance, shutdown margin, temperature coefficient and power distribution. Using a detector element that spans the length of the active fuel region inside the core provides a signal of total integrated flux. Measuring the integrated flux distributions and changes at various rodded conditions and plant temperatures, and comparing them to predicted flux levels, validates all core necessary core design characteristics. INCISV eliminates the dependence on various corrections and assumptions between the ex-core detectors and the core for traditional physics testing programs. This program also eliminates the need for special rod maneuvers which are infrequently performed by plant operators during typical core design verification testing and allows for safer startup activities. (authors)

Exploring the conformational and reactive dynamics of biomolecules in solution using an extended version of the glycine reactive force field

DEFF Research Database (Denmark)

Monti, Susanna; Corozzi, Alessandro; Fristrup, Peter

2013-01-01

In order to describe possible reaction mechanisms involving amino acids, and the evolution of the protonation state of amino acid side chains in solution, a reactive force field (ReaxFF-based description) for peptide and protein simulations has been developed as an expansion of the previously rep...

Analysis of criticality safety of coupled fast-thermal core 'HERBE'

International Nuclear Information System (INIS)

Pesic, M.

1991-01-01

Power excursion during possible fast core flooding is analyzed as serious accident. Model gives short filling time of fast zone with moderator after break of fast core tank. Reactivity increase is determined by computer codes and verified in specific experiments. Measurements of safety rods drop time and reactivity worth are performed. Coupled core kinetics parameters are determined according to model of Avery. Power excursion study, depending on power level threshold and safety instrumentation response time is performed. It was shown that safety system can shut-down reactor safely even in case of highly set power thresholds and partially failure of safety chain. (author)

Surface reactivity and layer analysis of chemisorbed reaction films in ...

Indian Academy of Sciences (India)

Administrator

Surface reactivity and layer analysis of chemisorbed reaction films in ... in the nitrogen environment. Keywords. Surface reactivity ... sium (Na–K) compounds in the coating or core of the ..... Barkshire I R, Pruton M and Smith G C 1995 Appl. Sur.

Evaluation of RSG-GAS Core Management Based on Burnup Calculation

International Nuclear Information System (INIS)

Lily Suparlina; Jati Susilo

2009-01-01

Evaluation of RSG-GAS Core Management Based on Burnup Calculation. Presently, U 3 Si 2 -Al dispersion fuel is used in RSG-GAS core and had passed the 60 th core. At the beginning of each cycle the 5/1 fuel reshuffling pattern is used. Since 52 nd core, operators did not use the core fuel management computer code provided by vendor for this activity. They use the manually calculation using excel software as the solving. To know the accuracy of the calculation, core calculation was carried out using two kinds of 2 dimension diffusion codes Batan-2DIFF and SRAC. The beginning of cycle burn-up fraction data were calculated start from 51 st to 60 th using Batan-EQUIL and SRAC COREBN. The analysis results showed that there is a disparity in reactivity values of the two calculation method. The 60 th core critical position resulted from Batan-2DIFF calculation provide the reduction of positive reactivity 1.84 % Δk/k, while the manually calculation results give the increase of positive reactivity 2.19 % Δk/k. The minimum shutdown margin for stuck rod condition for manual and Batan-3DIFF calculation are -3.35 % Δk/k dan -1.13 % Δk/k respectively, it means that both values met the safety criteria, i.e <-0.5 % Δk/k. Excel program can be used for burn-up calculation, but it is needed to provide core management code to reach higher accuracy. (author)

An analysis of reactivity prediction during the reactor start-up process

International Nuclear Information System (INIS)

Bajgl, Josef; Krysl, Vaclav; Svarny, Jiri

2015-01-01

The different VVER-440 core fuel loadings subcriticality evaluations are performed during the start-up process by boron dilution or control assembly withdrawn by macrocode MOBY-DICK calculations. The dynamic reactivity and quasicritical reactivity are compared and sensitivity of reactivity prediction at the low boundary of start-up interval (ρ = -0,01) has been provided on the basis of different modelling of ionization chamber (IC) response calculation. Special attention is paid to the impact of power distribution and spontaneous fission distribution form factor on IC response correction during control assembly movement. Precision and robustness of different corrections of IC signal processing in real core start-up processed IC signals was evaluated.

Development of a computer-aided digital reactivity computer system for PWRs

International Nuclear Information System (INIS)

Chung, S.-K.; Sung, K.-Y.; Kim, D.; Cho, D.-Y.

1993-01-01

Reactor physics tests at initial startup and after reloading are performed to verify nuclear design and to ensure safety operation. Two kinds of reactivity computers, analog and digital, have been widely used in the pressurized water reactor (PWR) core physics test. The test data of both reactivity computers are displayed only on the strip chart recorder, and these data are managed by hand so that the accuracy of the test results depends on operator expertise and experiences. This paper describes the development of the computer-aided digital reactivity computer system (DRCS), which is enhanced by system management software and an improved system for the application of the PWR core physics test

Parametric study of postulated reactivity transients due to ingress of heavy water from the reflector tank into the converted core of APSARA reactor

International Nuclear Information System (INIS)

Sankaranarayanan, S.

2004-01-01

Research reactors in the power range 5-10 MW with useable neutron flux values >1.OE+14 n/sqcm/sec can be constructed using LEU fuel with light water for neutron moderation and fuel cooling. In order to obtain a large irradiation volume, a heavy water reflector is used where fairly high neutron flux levels can be obtained. A prototype LEU fuelled 5/10 MW reactor design has been developed in the Bhabha Atomic Research Centre in Trombay. Work is on hand to carry out technology simulation of this reactor design by converting the pool type reactor APSARA in BARC. Presently the Apsara reactor uses MTh type high enriched U-Al alloy plate type fuel loaded in a 7x7 grid with a square lattice pitch of 76.8 mm. The reactor has three control-scram-shut off rods and one regulating control rod. In the first phase of the simulation studies, it is proposed to use the existing high enriched uranium fuel in a modified core with 37 positions arranged with a square lattice pitch of 84.8 mm, surrounded by a 50 cm thick heavy water reflector. Subsequently the converted core will use plate-type low enriched uranium suicide fuel. One of the accident scenarios postulated for the safety evaluation of the modified APSARA reactor is the reactivity transient due to the ingress of heavy water into the core through a small sized rupture in the aluminium wall of the reflector tank. Parametric analyses were done for the safety evaluation of modified Apsara reactor, for postulated leak of heavy water into the core from the reflector tank. A simplified computer code REDYN, based on point model reactor kinetics with one effective group of delayed neutrons is used for the analyses. Results of several parametric cases used in the study show that it is possible to contain the consequences of this type of reactivity transient within acceptable fuel and coolant thermal safety limits

Spatial Linkage and Urban Expansion: AN Urban Agglomeration View

Science.gov (United States)

Jiao, L. M.; Tang, X.; Liu, X. P.

2017-09-01

Urban expansion displays different characteristics in each period. From the perspective of the urban agglomeration, studying the spatial and temporal characteristics of urban expansion plays an important role in understanding the complex relationship between urban expansion and network structure of urban agglomeration. We analyze urban expansion in the Yangtze River Delta Urban Agglomeration (YRD) through accessibility to and spatial interaction intensity from core cities as well as accessibility of road network. Results show that: (1) Correlation between urban expansion intensity and spatial indicators such as location and space syntax variables is remarkable and positive, while it decreases after rapid expansion. (2) Urban expansion velocity displays a positive correlation with spatial indicators mentioned above in the first (1980-1990) and second (1990-2000) period. However, it exhibits a negative relationship in the third period (2000-2010), i.e., cities located in the periphery of urban agglomeration developing more quickly. Consequently, the hypothesis of convergence of urban expansion in rapid expansion stage is put forward. (3) Results of Zipf's law and Gibrat's law show urban expansion in YRD displays a convergent trend in rapid expansion stage, small and medium-sized cities growing faster. This study shows that spatial linkage plays an important but evolving role in urban expansion within the urban agglomeration. In addition, it serves as a reference to the planning of Yangtze River Delta Urban Agglomeration and regulation of urban expansion of other urban agglomerations.

Using nodal expansion method in calculation of reactor core with square fuel assemblies

International Nuclear Information System (INIS)

Abdollahzadeh, M. Y.; Boroushaki, M.

2009-01-01

A polynomial nodal method is developed to solve few-group neutron diffusion equations in cartesian geometry. In this article, the effective multiplication factor, group flux and power distribution based on the nodal polynomial expansion procedure is presented. In addition, by comparison of the results the superiority of nodal expansion method on finite-difference and finite-element are fully demonstrated. The comparison of the results obtained by these method with those of the well known benchmark problems have shown that they are in very good agreement.

Design study on PWR-type reduced-moderation light water core. Investigation of core adopting seed-blanket fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Shimada, Shoichiro; Kugo, Teruhiko; Okubo, Tsutomu; Iwamura, Takamichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2003-03-01

As a part of the design study on PWR-type Reduced-Moderation Water Reactors (RMWRs), a light water cooled core with the seed-blanket type fuel assemblies has been investigated. An assembly with seed of 13 layers and blanket of 5 layers was selected by optimization calculations. The core was composed with the 163 assemblies. The following results were obtained by burn-up calculations with the MVP-BURN code; The cycle length is 15 months by 3-batch refueling. The discharge burn-up including the inner blanket is about 25 GWd/t. The conversion ratio is about 1.0. The void reactivity coefficient is about-26.1 pcm/%void at BOC and -21.7pcm%void at EOC. About 10% of MA makes conversion ratio decrease about 0.05 to obtain the same burn-up. The void reactivity coefficient increased significantly and it is necessary to reduce it. FP amount corresponding to about 2 % of total plutonium weight makes reactivity decrease about 0.5 %{delta}k/k and void reactivity coefficient increase, however these changes are within the design margins. Capability of multi-recycling of plutonium was confirmed, using discharged plutonium for 4 cycles, if fissile plutonium of 15.5wt% is used. The conversion ratio increases by about 0.026 with recycling. However, void reactivity coefficient increases and some effort to obtain negative void reactivity coefficient is necessary. (author)

Scoping Analysis on Core Disruptive Accident in PGSFR (2015 Results)

Energy Technology Data Exchange (ETDEWEB)

Lee, Seung Won; Chang, Won-Pyo; Ha, Kwi-Seok; Ahn, Sang June; Kang, Seok Hun; Choi, Chi-Woong; Lee, Kwi Lim; Jeong, Jae-Ho; Kim, Jin Su; Jeong, Taekyeong [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In general, the severe accident is classified by three phases. The first phase is the initiation (pre-disassembly) phase that occurs the gradual core meltdown from accident initiation to the point of neutronic shutdown with an intact geometry. The second phase is the transition phase that happens the fuel transition from a solid to a liquid phase. Fuel and cladding can melt to form a molten pool and core can boil, then criticality conditions can recur. The third phase is the disassembly phase. In other words, this phase is Core Disruptive Accident (CDA). Power excursion is followed until the core is disassembled in this phase. In the early considerations of Liquid Metal Fast Breeder Reactor (LMFBR) energetics, the term Hypothetical Core Disruptive Accidents (HCDAs) was in common use. This was not only to connote the extremely low probability of initiation of such accidents, but also the tentative nature of our understanding of their behavior and resulting consequences. A numerical analysis is conducted to estimate the energy release, pressure behavior and core expansion behavior induced by CDA of PGSFR using CDA-ER and CDA-CEME codes. Conservatively, the calculated results of energy release and pressure behavior induced by CDA without Doppler effect in PGSFR when whole cores were melted (100 $/s) were 7.844 GJ and 4.845 GPa, respectively. With Doppler effect, the analyzed maximum energy release and pressure were 6.696 GJ and 3.449 GPa, respectively. The calculated results of the core expansion behavior during 0.015 seconds after the explosion without Doppler effect in PGSFR when whole cores were melted (100 $/s) were as follows: The total energy is calculated to be 1.87 GJ. At 0.01 s, the kinetic energy of the sodium is 1.85 GJ, while the expansion work and internal energy of the bubble are 19.7 MJ and 0.98 J, respectively. With Doppler effect, the total energy is calculated to be 1.33 GJ. At 0.01 s, the kinetic energy of the sodium is 1.31 GJ, while the expansion

Calculation of the RSG-GAS core using computer code citation-3D

International Nuclear Information System (INIS)

Taryo, T.; Rokhmadi

1998-01-01

Since core reactivity is one of the reactor safety parameters, this R and D has been carried out. To carry out the R and D, the code called WIMSD4 was used respectively for generating cross section and diffusion parameters. The code CITATION was then applied to estimate core reactivity in the RSG-GAS core. To verify the result of the calculation, data and information of the RSG-GAS Typical Working Core Were used. To Prove the codes reliably used, the case of all control elements down in the reactor core and that of all control rods up in the core were applied. The result taking into account those cases showed respectively that K eff are less and greater than unity (K eff eff >1)

Core design for use with precision composite reflectors

Science.gov (United States)

Porter, Christopher C. (Inventor); Jacoy, Paul J. (Inventor); Schmitigal, Wesley P. (Inventor)

1992-01-01

A uniformly flexible core, and method for manufacturing the same, is disclosed for use between the face plates of a sandwich structure. The core is made of a plurality of thin corrugated strips, the corrugations being defined by a plurality of peaks and valleys connected to one another by a plurality of diagonal risers. The corrugated strips are orthogonally criss-crossed to form the core. The core is particularly suitable for use with high accuracy spherically curved sandwich structures because undesirable stresses in the curved face plates are minimized due to the uniform flexibility characteristics of the core in both the X and Y directions. The core is self venting because of the open geometry of the corrugations. The core can be made from any suitable composite, metal, or polymer. Thermal expansion problems in sandwich structures may be minimized by making the core from the same composite materials that are selected in the manufacture of the curved face plates because of their low coefficients of thermal expansion. Where the strips are made of a composite material, the core may be constructed by first cutting an already cured corrugated sheet into a plurality of corrugated strips and then secondarily bonding the strips to one another or, alternatively, by lying a plurality of uncured strips orthogonally over one another in a suitable jig and then curing and bonding the entire plurality of strips to one another in a single operation.

The expansion of electric power systems considering the operation reliability; Expansion de sistemas electricos considerando la securidad de servicio

Energy Technology Data Exchange (ETDEWEB)

Cortes, Marcelo A. [Universidad de Antofagasta (Chile). Dept. de Ingenieria Electrica; Moya A, Oscar [Chile Univ., Santiago (Chile). Dept. de Ingenieria Civil

1997-12-31

This publication describes a procedure based in an linear programming for the planning of the expansion of mediating term in power systems. The implicit alternatives for network expansion, whether power plants or transmission lines, are classified according to indexes of reinforcement. The indexes are gotten resulting shadow pricing upon resolving a problem of minimum cost of operation and outage. The network is subjected to the most probable contingencies and from these outputs the expected reinforcement results are obtained. The problem of the reactive power accounts for restrictions in line capacity and prior evaluation of A C power flux. Results obtained are used for methodology validation. (author) 12 refs., 1 fig., 3 tabs.; e-mail: mcortes at apolo.cc.uantof.cl.; osmoya at tamarugo.cec.uchile.cl

HALO EXPANSION IN COSMOLOGICAL HYDRO SIMULATIONS: TOWARD A BARYONIC SOLUTION OF THE CUSP/CORE PROBLEM IN MASSIVE SPIRALS

Energy Technology Data Exchange (ETDEWEB)

Maccio, A. V.; Stinson, G. [Max-Planck-Institut fuer Astronomie, 69117 Heidelberg (Germany); Brook, C. B.; Gibson, B. K. [University of Central Lancashire, Jeremiah Horrocks Institute for Astrophysics and Supercomputing, Preston PR1 2HE (United Kingdom); Wadsley, J.; Couchman, H. M. P. [Department of Physics and Astronomy, McMaster University, Hamilton, Ontario, L8S 4M1 (Canada); Shen, S. [Department of Astronomy and Astrophysics, University of California Santa Cruz, Santa Cruz, CA 95064 (United States); Quinn, T., E-mail: maccio@mpia.de, E-mail: stinson@mpia.de [Astronomy Department, University of Washington, Seattle, WA 98195-1580 (United States)

2012-01-15

A clear prediction of the cold dark matter (CDM) model is the existence of cuspy dark matter halo density profiles on all mass scales. This is not in agreement with the observed rotation curves of spiral galaxies, challenging on small scales the otherwise successful CDM paradigm. In this work we employ high-resolution cosmological hydrodynamical simulations to study the effects of dissipative processes on the inner distribution of dark matter in Milky Way like objects (M Almost-Equal-To 10{sup 12} M{sub Sun }). Our simulations include supernova feedback, and the effects of the radiation pressure of massive stars before they explode as supernovae. The increased stellar feedback results in the expansion of the dark matter halo instead of contraction with respect to N-body simulations. Baryons are able to erase the dark matter cuspy distribution, creating a flat, cored, dark matter density profile in the central several kiloparsecs of a massive Milky-Way-like halo. The profile is well fit by a Burkert profile, with fitting parameters consistent with the observations. In addition, we obtain flat rotation curves as well as extended, exponential stellar disk profiles. While the stellar disk we obtain is still partially too thick to resemble the Milky Way thin disk, this pilot study shows that there is enough energy available in the baryonic component to alter the dark matter distribution even in massive disk galaxies, providing a possible solution to the long-standing problem of cusps versus cores.

Evaluation of core distortion in FBR

International Nuclear Information System (INIS)

Ikarimoto, I.; Tanaka, M.; Okubo, Y.

1984-01-01

The analyses of FBR's core distortion are mainly performed in order to evaluate the following items: 1) Change of reactivity; 2) Force at pads on core assemblies; 3) Withdrawal force at refueling; 4) Loading, refueling and residual deviations of wrapper tubes (core assemblies) at the top; 5) Bowing modes of guide tubes for control rods. The analysis of core distortion are performed by using computer program for two-dimensional row deformation analysis or three-dimensional core deformation if necessary, considering these evaluated items which become design conditions. This report shows the relationship between core deformation analysis and component design, a point of view of choosing an analysis program for design considering core characteristics, and computing examples of core deformation of prototype class reactor by the above code. (author)

Structural and reactivity models for copper oxygenases: cooperative effects and novel reactivities.

Science.gov (United States)

Serrano-Plana, Joan; Garcia-Bosch, Isaac; Company, Anna; Costas, Miquel

2015-08-18

Dioxygen is widely used in nature as oxidant. Nature itself has served as inspiration to use O2 in chemical synthesis. However, the use of dioxygen as an oxidant is not straightforward. Its triplet ground-state electronic structure makes it unreactive toward most organic substrates. In natural systems, metalloenzymes activate O2 by reducing it to more reactive peroxide (O2(2-)) or superoxide (O2(-)) forms. Over the years, the development of model systems containing transition metals has become a convenient tool for unravelling O2-activation mechanistic aspects and reproducing the oxidative activity of enzymes. Several copper-based systems have been developed within this area. Tyrosinase is a copper-based O2-activating enzyme, whose structure and reactivity have been widely studied, and that serves as a paradigm for O2 activation at a dimetal site. It contains a dicopper center in its active site, and it catalyzes the regioselective ortho-hydroxylation of phenols to catechols and further oxidation to quinones. This represents an important step in melanin biosynthesis and it is mediated by a dicopper(II) side-on peroxo intermediate species. In the present accounts, our research in the field of copper models for oxygen activation is collected. We have developed m-xylyl linked dicopper systems that mimick structural and reactivity aspects of tyrosinase. Synergistic cooperation of the two copper(I) centers results in O2 binding and formation of bis(μ-oxo)dicopper(III) cores. These in turn bind and ortho-hydroxylate phenolates via an electrophilic attack of the oxo ligand over the arene. Interestingly the bis(μ-oxo)dicopper(III) cores can also engage in ortho-hydroxylation-defluorination of deprotonated 2-fluorophenols, substrates that are well-known enzyme inhibitors. Analysis of Cu2O2 species with different binding modes show that only the bis(μ-oxo)dicopper(III) cores can mediate the reaction. Finally, the use of unsymmetric systems for oxygen activation is a field

Reference core design Mark-I and -II of the experimental, multi-purpose, high-temperature, gas-cooled reactor

International Nuclear Information System (INIS)

Shindo, Ryuiti; Hirano, Mitsumasa; Aruga, Takeo; Yasukawa, Sigeru

1977-10-01

Reactivity worth of the control rods and power distribution in the initial hot-clean core of reference core design Mark-I and -II have been studied. The need for burnable poison was confirmed, because of the limitations in number, diameter and reactivity worth of the control rods due to structures of pressure vessel and fuel element and to safety of the core. While the initial excess reactivity is reduced by use of the burnable poison, the recovery of core reactivity with burnup of the burnable poison requires a complicated withdrawal sequence of the control rods. The radial power gradient in the core is not large, due to orifice control of the coolant helium flow, effectiveness of the reflector in the small core and continuous distribution of burnup in the core by one-batch refuelling scheme. The local peaking factor in unit orifice regions, therefore, is the most important core design. Control of the axial power distribution is necessary to reduce the maximum fuel temperature and the exponential power distribution peaked toward the inlet of the core is most suitable. However, insertion of the control rods from top of the core disturbs the axial power distribution, so this effect must be considered in design of the withdrawal sequence of control rods. Nuclear properties of the core were revealed from results of the study for the initial hot-clean core. (auth.)

The exponential function expansion of the intra-nodal cross sections for the spectral history gradient correction

International Nuclear Information System (INIS)

Cho, J. Y.; Noh, J. M.; Cheong, H. K.; Choo, H. K.

1998-01-01

In order to simplify the previous spectral history effect correction based on the polynomial expansion nodal method, a new spectral history effect correction is proposed. The new spectral history correction eliminates four microscopic depletion points out of total 13 depletion points in the previous correction by approximating the group cross sections with exponential function. The neutron flux to homogenize the group cross sections for the correction of the spectral history effect is calculated by the analytic function expansion nodal method in stead of the conventional polynomial expansion nodal method. This spectral history correction model is verified against the three MOX benchmark cores: a checkerboard type, a small core with 25 fuel assemblies, and a large core with 177 fuel assemblies. The benchmark results prove that this new spectral history correction model is superior to the previous one even with the reduced number of the local microscopic depletion points

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

International Nuclear Information System (INIS)

Jernkvist, Lars Olof; Massih, Ali R.

2004-12-01

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

PBDOWN: A computer code for simulation of core material discharge and expansion in the upper coolant plenum in a hypothetical unprotected loss of flow accident in a LMFBR

International Nuclear Information System (INIS)

Royl, P.

1985-01-01

The report gives a description of the code PBDOWN (Pool Blow Down), its equations, input specifications and subroutines and it lists the input and output for some samples. Besides that some analysis results for the SNR-300 are discussed, that were obtained with this code. PBDOWN is an integral blow-down and expansion code, which simulates core material discharge and expansion into a sodium filled upper coolant plenum after build-up of vapour pressures in an unprotected loss of flow accident. The model includes the effect of sodium entrainment into an expending bubble of fuel or steel vapour with various assumptions for the heat transfer and vaporization of the entrained sodium droplets. The expanding vapour bubble is connected to the discharging pool via an orifice of a given size through which a time dependent ejection is simulated using quasi-stationary blow down correlations. The model allows bounding analysis of the possible influence of sodium vapour as a secondary working fluid, that is activated outside the pool on the overall expansion energy and discharge

Studies on the inhomogeneous core density of a fluidized bed nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Van der Hagen, T.H.J.J.; Van Dam, H.; Hoogenboom, J.E.; Khotylev, V.A. [Delft Univ. of Technology (Netherlands). Interfaculty Reactor Inst.; Harteveld, W.; Mudde, R.F.

1997-12-31

Results are reported on the expected time dependent core density profile of a fluidized-bed nuclear fission reactor. Core densities have been measured in a test facility by the gamma-transmission technique. Bubble and particle-cluster sizes, positions, velocities and frequencies could be determined. Neutronic studies have been performed on the influence of core voids on reactivity using Monte-Carlo and neutron-transport codes. Fuel-particle importance has been determined. Point-kinetic parameters have been calculated for linking reactivity perturbations to power fluctuations. (author)

Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

Energy Technology Data Exchange (ETDEWEB)

Won, Jong Hyuck; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr; Park, Hae Min; Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr

2014-12-15

Highlights: • The graphite rod-inserted SFR fuel assembly is proposed to achieve low sodium void reactivity. • The neutronics/thermal-hydraulics analyses are performed for the proposed SFR cores. • The sodium void reactivity is improved about 960–1030 pcm compared to reference design. - Abstract: The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B{sub 4}C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes. In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030 pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly. The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

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.

Joyo ATWS test analysis by Mimir-N2

International Nuclear Information System (INIS)

Yoshida, Akihiro

2001-03-01

The study on the passive safety test by using the Experimental Fast Reactor Joyo was performed to demonstrate the inherent safety of fast breeder reactors. An analysis code: Mimir-N2, which has been developed to analyze Joyo plant kinetics, was selected as a standard code for this study. In order to increase the reliability of the calculation, Mimir-N2 code was adjusted based on the data obtained through several plant characteristics tests carried out in Joyo. Throughout an operational data obtained in Joyo, it is supposed that the burn-up dependency observed on the power reactivity coefficient might be coming from the reactivity shift caused by a depression of a thermal expansion of fuel pellet. Based on the relationship between the measured power reactivity coefficient and the core averaged burn-up, the burn-up dependency mentioned above was estimated and introduced to Mimir-N2. As a result, calculated core and plant dynamics during the step reactivity response test, such as the response of the power range neutron monitor and the coolant temperature at the core inlet/outlet, corresponded with the measured value. Especially, it was confirmed that Mimir-N2 can simulate the perturbation caused by the thermal expansion of the core support plate. In addition, Mimir-N2 was modified to be enable to take into account for the core bowing reactivity, which is calculated by the core bowing reactivity analysis system developed for Joyo. The preliminary analysis of the plant dynamics during the ATWS events in MK-III core were carried out by using modified Mimir-N2. As a result, it was confirmed that the core bowing reactivity should not be neglected because it sometimes shows positive feedback characteristics. (author)

Contribution to the qualification of calculation methods of reactivity and of flux and power distributions in nuclear pressurized water reactor cores

International Nuclear Information System (INIS)

Abit, K.

1984-01-01

The last stage of the creation computer methods and calculations consists of verifying the running and qualifying the results obtained. The work of the present thesis consisted of improving a coupling method between radial and axial phenomena in a PWR core, refering to three-dimensional calculations, while ensuring a perfect coherence between the programmed physical models. The calculation results have been compared to measurements of reactivity and of flux distributions realized during start-up tests. Thus, the methods have been applied to the calculation of the evolution of a burnable poison (gadolinium) in view of operation in long campaign. 13 refs [fr

An axially and radially two-zoned large liquid-metal fast breeder reactor core concept

International Nuclear Information System (INIS)

Kamei, T.; Arie, K.; Moriki, Y.; Suzuki, M.; Yamaoka, M.

1985-01-01

A new core concept that has advantages over conventional homogeneous cores in neutronics characteristics such as power peaking factor, burnup reactivity loss, and reactivity response to the movement of control rods in earthquakes has been evolved. Two options of the new core concept are feasible. One is the so-called axially heterogeneous core, with the internal blanket placed at the lower part of the core. The other concept is similar to the conventional homogeneous core, but has two different plutonium-enriched zones in the axial as well as in the radial direction, so it is a hybrid type of the conventional homogeneous core and the axially heterogeneous core. The new design concept is described and the way that the core characteristics are improved by the chosen key parameters is shown

Multi-site reactivity: reduction of six equivalents of nitrite to give an Fe6(NO)6 cluster with a dramatically expanded octahedral core.

Science.gov (United States)

Harris, T David; Betley, Theodore A

2011-09-07

Reaction of NO(2)(-) with the octahedral cluster ((H)L)(2)Fe(6) in the presence of a proton source affords the hexanitrosyl cluster ((H)L)(2)Fe(6)(NO)(6). This species forms via a proton-induced reduction of six nitrite molecules per cluster, utilizing each site available on the polynuclear core. Formation of the hexanitrosyl cluster is accompanied by a near 2-fold expansion of the ((H)L)(2)Fe(6) core volume, where intracore Fe-Fe interactions are overcome by strong π-bonding between Fe centers and NO ligands. A core volume of this magnitude is rare in octahedral metal clusters not supported by interstitial atoms. Moreover, the structural flexibility afforded by the ((H)L)(2)Fe(6) platform highlights the potential for other reaction chemistry involving species with metal-ligand multiple bonds. Carrying out the reaction of the cluster [((H)L)(2)Fe(6)(NCMe)(6)](4+) with nitrite in the absence of a proton source serves to forestall the nitrite reduction and enables clean isolation of the intermediate hexanitro cluster [((H)L)(2)Fe(6)(NO(2))(6)](2-).

Temperature coefficients of reactivity in the fourth loading of ZENITH

Energy Technology Data Exchange (ETDEWEB)

Caro Manso, R; Freemantle, R G; Rogers, J D [Graphite Reactor Physics Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

1962-10-15

Measurements have been made of the temperature coefficients of reactivity associated with the core plus end reflectors and the side reflector of the fourth core loading of ZENITH, which had a carbon/U235 atomic ratio of 7788 and no other absorber. (author)

Temperature coefficients of reactivity in the fourth loading of ZENITH

International Nuclear Information System (INIS)

Caro Manso, R.; Freemantle, R.G.; Rogers, J.D.

1962-10-01

Measurements have been made of the temperature coefficients of reactivity associated with the core plus end reflectors and the side reflector of the fourth core loading of ZENITH, which had a carbon/U235 atomic ratio of 7788 and no other absorber. (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)

Improvement of Cycle Dependent Core Model for NPP Simulator

International Nuclear Information System (INIS)

Song, J. S.; Koo, B. S.; Kim, H. Y. and others

2003-11-01

The purpose of this study is to establish automatic core model generation system and to develop 4 cycle real time core analysis methodology with 5% power distribution and 500 pcm reactivity difference criteria for nuclear power plant simulator. The standardized procedure to generate database from ROCS and ANC, which are used for domestic PWR core design, was established for the cycle specific simulator core model generation. An automatic data interface system to generate core model also established. The system includes ARCADIS which edits group constant and DHCGEN which generates interface coupling coefficient correction database. The interface coupling coefficient correction method developed in this study has 4 cycle real time capability and accuracies of which the maximum differences between core design results are within 103 pcm reactivity, 1% relative power distribution and 6% control rod worth. A nuclear power plant core simulation program R-MASTER was developed using the methodology and applied by the concept of distributed client system in simulator. The performance was verified by site acceptance test in Simulator no. 2 in Kori Training Center for 30 initial condition generation and 27 steady state, transient and postulated accident situations

Improvement of Cycle Dependent Core Model for NPP Simulator

Energy Technology Data Exchange (ETDEWEB)

Song, J. S.; Koo, B. S.; Kim, H. Y. and others

2003-11-15

The purpose of this study is to establish automatic core model generation system and to develop 4 cycle real time core analysis methodology with 5% power distribution and 500 pcm reactivity difference criteria for nuclear power plant simulator. The standardized procedure to generate database from ROCS and ANC, which are used for domestic PWR core design, was established for the cycle specific simulator core model generation. An automatic data interface system to generate core model also established. The system includes ARCADIS which edits group constant and DHCGEN which generates interface coupling coefficient correction database. The interface coupling coefficient correction method developed in this study has 4 cycle real time capability and accuracies of which the maximum differences between core design results are within 103 pcm reactivity, 1% relative power distribution and 6% control rod worth. A nuclear power plant core simulation program R-MASTER was developed using the methodology and applied by the concept of distributed client system in simulator. The performance was verified by site acceptance test in Simulator no. 2 in Kori Training Center for 30 initial condition generation and 27 steady state, transient and postulated accident situations.

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)

Series expansion of two-dimensional fields produced by iron-core magnets

International Nuclear Information System (INIS)

Satoh, Kotaro.

1997-02-01

This paper discusses the validity of a series expansion of two-dimensional magnetic fields with harmonic functions, and suggests that the series may not converge outside of the pole gap. It also points out that this difficulty may appear due to a slow convergence of the series near to the pole edge, even within the convergent area. (author)

Correlation and flux tilt measurements of coupled-core reactor assemblies

International Nuclear Information System (INIS)

Harries, J.R.

1976-01-01

The systematics of coupling reactivity and time delay between cores have been investigated with a series of coupled-core assemblies on the AAEC Split-table Critical Facility. The assemblies were similar to the Universities' Training Reactor (UTR), but had graphite coupling region thickness of 450 mm, 600 mm and 800 mm. The coupling reactivity measured by both the cross-correlation of reactor noise and the flux tilt methods was stronger than for the UTRs, but showed a similar trend with core spacing. The cross-correlograms were analysed using the two-node model to derive the time delays between the cores. The time delays were compared with thermal neutron wave propagation, and found to be consistent when the time delays were added to the individual node response-function delays. (author)

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)

Design of 50 MWe HTR-PBMR reactor core and nuclear power plant fuel using SRAC2006 programme

International Nuclear Information System (INIS)

Bima Caraka Putra; Yosaphat Sumardi; Yohannes Sardjono

2014-01-01

This research aims to assess the design of core and fuel of nuclear power plant type High Temperature Reactor-Pebble Bed Modular Reactor 50 MWe from the Beginning of Life (BOL) to Ending of life (EOL) with eight years operating life. The parameters that need to be analyzed in this research are the temperature distribution inside the core, quantity enrichment of U 235 , fuel composition, criticality, and temperature reactivity coefficient of the core. The research was conducted with a data set of core design parameters such as nuclides density, core and fuel dimensions, and the axial temperature distribution inside the core. Using SRAC2006 program package, the effective multiplication factor (k eff ) values obtained from the input data that has been prepared. The results show the value of the criticality of core is proportional to the addition of U 235 enrichment. The optimum enrichment obtained at 10.125% without the use of burnable poison with an excess reactivity of 3.1 2% at BOL. The addition Gd 2O3 obtained an optimum value of 12 ppm burnable poison with an excess reactivity 0.38 %. The use of Er 2O3 with an optimum value 290 ppm has an excess reactivity 1.24 % at BOL. The core temperature reactivity coefficient with and without the use of burnable poison has a negative values that indicates the nature of its inherent safety. (author)

Sensitivity Tests for the Unprotected Events of the Prototype Gen-IV SFR

Energy Technology Data Exchange (ETDEWEB)

Choi, Chiwoong; Lee, Kwilim; Jeong, Jaeho; Yu, Jin; An, Sangjun; Lee, Seung Won; Chang, Wonpyo; Ha, Kwiseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

Unprotected Transient Over Power, (UTOP), Unprotected Loss Of Flow (ULOF), and Unprotected Loss Of Heat Sink (ULOHS) are selected as ATWS events. Among these accidents, the ULOF event shows the lowest clad temperature. However, the ULOHS event showed the highest peak clad temperature, due to the positive CRDL/RV expansion reactivity feedback and insufficient DHRS capacity. In this study, the sensitivity tests are conducted. In the case of the UTOP event, a sensitivity test for the reactivity insertion amount and rate were conducted. This analysis can give a requirement for margin of control rod stop system (CRSS). Currently, the reactivity feedback model for the PGSFR is not validated yet. However, the reactivity feedback models in the MARS-LMR are validating with various plant-based data including EBR-II SHRT. The ATWS events for the PGSFR classified in the design extended condition including UTOP, ULOF, and ULOHS are analyzed with MARS-LMR. In this study, the sensitivity tests for reactivity insertion amount and rate in the UTOP event are conducted. The reactivity insertion amount is obviously an influential parameter. The reactivity insertion amount can give a requirement for design of the CRSS, therefore, this sensitivity result is very important to the CRSS. In addition, sensitivity tests for the weighting factor in the radial expansion reactivity model are carried out. The weighting factor for a grid plate, W{sub GP}, which means contribution of feedback in the grid plate is changed for all unprotected events. The grid plate expansion is governed by a core inlet temperature. As the W{sub GP} is increased, the power in the UTOP and the ULOF is increased, however, the power in the ULOHS is decreased. The higher power during transient means lower reactivity feedback and smaller expansion. Thus, the core outlet temperature rise is dominant in the UTOP and ULOF events, however, the core inlet temperature rise is dominant in the ULOHS. Therefore, the grid plate

Development of a computer code for neutronic calculations of a hexagonal lattice of nuclear reactor using the flux expansion nodal method

Directory of Open Access Journals (Sweden)

Mohammadnia Meysam

2013-01-01

Full Text Available The flux expansion nodal method is a suitable method for considering nodalization effects in node corners. In this paper we used this method to solve the intra-nodal flux analytically. Then, a computer code, named MA.CODE, was developed using the C# programming language. The code is capable of reactor core calculations for hexagonal geometries in two energy groups and three dimensions. The MA.CODE imports two group constants from the WIMS code and calculates the effective multiplication factor, thermal and fast neutron flux in three dimensions, power density, reactivity, and the power peaking factor of each fuel assembly. Some of the code's merits are low calculation time and a user friendly interface. MA.CODE results showed good agreement with IAEA benchmarks, i. e. AER-FCM-101 and AER-FCM-001.

Calculation of the Reactivity Equivalence of Control Rods in the Second Charge of the HBWR

International Nuclear Information System (INIS)

Weissglas, P.

1960-11-01

Full text: Using current methods the reactivity equivalence of 19 31 and 37 centrally located control rods in the second charge of the HBWR has been calculated. An estimate of the available excess reactivity with clean cold core has also been made. Insertion depth was taken as 0, l/3, 2/3 and 3/3 of the core length

Calculation of the Reactivity Equivalence of Control Rods in the Second Charge of the HBWR.

Energy Technology Data Exchange (ETDEWEB)

Weissglas, P [The Swedish State Power Board, Stockholm (Sweden)

1960-11-15

Full text: Using current methods the reactivity equivalence of 19 31 and 37 centrally located control rods in the second charge of the HBWR has been calculated. An estimate of the available excess reactivity with clean cold core has also been made. Insertion depth was taken as 0, l/3, 2/3 and 3/3 of the core length.

Whole core burnup calculations using `MCNP`

Energy Technology Data Exchange (ETDEWEB)

Haran, O; Shaham, Y [Israel Atomic Energy Commission, Beersheba (Israel). Nuclear Research Center-Negev

1996-12-01

Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It should be noted that burnup using Monte Carlo has been reported in the literatures, but this work is the result of an independent effort (authors).

Whole core burnup calculations using 'MCNP'

International Nuclear Information System (INIS)

Haran, O.; Shaham, Y.

1996-01-01

Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It should be noted that burnup using Monte Carlo has been reported in the literatures, but this work is the result of an independent effort (authors)

Space dependence of reactivity parameters on reactor dynamic perturbation measurements

International Nuclear Information System (INIS)

Maletti, R.; Ziegenbein, D.

1985-01-01

Practical application of reactor-dynamic perturbation measurements for on-power determination of differential reactivity weight of control rods and power coefficients of reactivity has shown a significant dependence of parameters on the position of outcore detectors. The space dependence of neutron flux signal in the core of a VVER-440-type reactor was measured by means of 60 self-powered neutron detectors. The greatest neutron flux alterations are located close to moved control rods and in height of the perturbation position. By means of computations, detector positions can be found in the core in which the one-point model is almost valid. (author)

Evolution of density-dependent movement during experimental range expansions.

Science.gov (United States)

Fronhofer, E A; Gut, S; Altermatt, F

2017-12-01

Range expansions and biological invasions are prime examples of transient processes that are likely impacted by rapid evolutionary changes. As a spatial process, range expansions are driven by dispersal and movement behaviour. Although it is widely accepted that dispersal and movement may be context-dependent, for instance density-dependent, and best represented by reaction norms, the evolution of density-dependent movement during range expansions has received little experimental attention. We therefore tested current theory predicting the evolution of increased movement at low densities at range margins using highly replicated and controlled range expansion experiments across multiple genotypes of the protist model system Tetrahymena thermophila. Although rare, we found evolutionary changes during range expansions even in the absence of initial standing genetic variation. Range expansions led to the evolution of negatively density-dependent movement at range margins. In addition, we report the evolution of increased intrastrain competitive ability and concurrently decreased population growth rates in range cores. Our findings highlight the importance of understanding movement and dispersal as evolving reaction norms and plastic life-history traits of central relevance for range expansions, biological invasions and the dynamics of spatially structured systems in general. © 2017 European Society For Evolutionary Biology. Journal of Evolutionary Biology © 2017 European Society For Evolutionary Biology.

Analysis on void reactivity of DCA lattice

International Nuclear Information System (INIS)

Min, B. J.; Noh, K. H.; Choi, H. B.; Yang, M. K.

2001-01-01

In case of loss of coolant accident, the void reactivity of CANDU fuel provides the positive reactivity and increases the reactor power rapidly. Therefore, it is required to secure credibility of the void reactivity for the design and analysis of reactor, which motivated a study to assess the measurement data of void reactivity. The assessment of lattice code was performed with the experimental data of void reactivity at 30, 70, 87 and 100% of void fractions. The infinite multiplication factors increased in four types of fuels as the void fractions of them grow. The infinite multiplication factors of uranium fuels are almost within 1%, but those of Pu fuels are over 10% by the results of WIMS-AECL and MCNP-4B codes. Moreover, coolant void reactivity of the core loaded with plutonium fuel is more negative compared with that with uranium fuel because of spectrum hardening resulting from large void fraction

A Reactive and Cycle-True IP Emulator for MPSoC Exploration

DEFF Research Database (Denmark)

Mahadevan, Shankar; Angiolini, Federico; Sparsø, Jens

2008-01-01

The design of MultiProcessor Systems-on-Chip (MPSoC) emphasizes intellectual-property (IP)-based communication-centric approaches. Therefore, for the optimization of the MPSoC interconnect, the designer must develop traffic models that realistically capture the application behavior as executing...... on the IP core. In this paper, we introduce a Reactive IP Emulator (RIPE) that enables an effective emulation of the IP-core behavior in multiple environments, including bit and cycle-true simulation. The RIPE is built as a multithreaded abstract instruction-set processor, and it can generate reactive...

Full MOX core for PWRs

International Nuclear Information System (INIS)

Puill, A.; Aniel-Buchheit, S.

1997-01-01

Plutonium management is a major problem of the back end of the fuel cycle. Fabrication costs must be reduced and plant operation simplified. The design of a full MOX PWR core would enable the number of reactors devoted to plutonium recycling to be reduced and fuel zoning to be eliminated. This paper is a contribution to the feasibility studies for achieving such a core without fundamental modification of the current design. In view of the differences observed between uranium and plutonium characteristics it seems necessary to reconsider the safety of a MOX-fuelled PWR. Reduction of the control worth and modification of the moderator density coefficient are the main consequences of using MOX fuel in a PWR. The core reactivity change during a draining or a cooling is thus of prime interest. The study of core global draining leads to the following conclusion: only plutonium fuels of very poor quality (i.e. with low fissile content) cannot be used in a 900 MWe PWR because of a positive global voiding reactivity effect. During a cooling accident, like an spurious opening of a secondary-side valve, the hypothetical return to criticality of a 100% MOX core controlled by means of 57 control rod clusters (made of hafnium-clad B 4 C rods with a 90% 10 B content) depends on the isotopic plutonium composition. But safety criteria can be complied with for all isotopic compositions provided the 10 B content of the soluble boron is increased to a value of 40%. Core global draining and cooling accidents do not present any major obstacle to the feasibility of a 100% MOX PWR, only minor hardware modifications will be required. (author)

Neutronic Core Performance of CAREM-25 Reactor

International Nuclear Information System (INIS)

Villarino, Eduardo; Hergenreder, Daniel; Matzkin, S

2000-01-01

The actual design state of core of CAREM-25 reactor is presented.It is shown that the core design complains with the safety and operation established requirements.It is analyzed the behavior of the reactor safety and control systems (single failure of the fast shut down system, single failure of the shut down system, single failure of the second shut down system, reactivity worth of the adjust and control system in normal operation and hot shut down, reactivity worth of the adjust and control system and the scheme of movement of the control rod during the operation cycle).It is shown the burnup profile of fuel elements with the proposed scheme of refueling and the burnup and power density distribution at different moments of the operation cycle.The power peaking factor of the equilibrium core is 2.56, the minimum DNBR is 1.90 and its average is 2.09 during the operation cycle

Reactivity Coefficient Calculation for AP1000 Reactor Using the NODAL3 Code

Science.gov (United States)

Pinem, Surian; Malem Sembiring, Tagor; Tukiran; Deswandri; Sunaryo, Geni Rina

2018-02-01

The reactivity coefficient is a very important parameter for inherent safety and stability of nuclear reactors operation. To provide the safety analysis of the reactor, the calculation of changes in reactivity caused by temperature is necessary because it is related to the reactor operation. In this paper, the temperature reactivity coefficients of fuel and moderator of the AP1000 core are calculated, as well as the moderator density and boron concentration. All of these coefficients are calculated at the hot full power condition (HFP). All neutron diffusion constant as a function of temperature, water density and boron concentration were generated by the SRAC2006 code. The core calculations for determination of the reactivity coefficient parameter are done by using NODAL3 code. The calculation results show that the fuel temperature, moderator temperature and boron reactivity coefficients are in the range between -2.613 pcm/°C to -4.657pcm/°C, -1.00518 pcm/°C to 1.00649 pcm/°C and -9.11361 pcm/ppm to -8.0751 pcm/ppm, respectively. For the water density reactivity coefficients, the positive reactivity occurs at the water temperature less than 190 °C. The calculation results show that the reactivity coefficients are accurate because the results have a very good agreement with the design value.

Safety characteristics of mid-sized MOX fueled liquid metal reactor core of high converter type in the initiating phase of unprotected loss of flow accident. Effect of low specific fuel power density on ULOF behavior brought by employment of large diameter fuel pins

International Nuclear Information System (INIS)

Ishida, Masayoshi; Kawada, Kenichi; Niwa, Hajime

2003-07-01

Safety characteristics in core disruptive accidents (CDAs) of mid-sized MOX fueled liquid metal reactor core of high converter type have been examined by using the CDA initiating phase analysis code SAS4A. The design concept of high converter type reactor core has been studied as one of options in the category of sodium-cooled reactor in Phase II of Feasibility Study on Commercialized Fast Reactor Cycle System. An unprotected loss-of-flow accident (ULOF) has been selected as a representative CDA initiator for this study. A core concept of high converter type, which employed a large diameter fuel pin of 11.1 mm with 1.2 m core height to get a large fuel volume fraction in the core to achieve high internal conversion ratio was proposed in JFY2001. Each fuel subassembly of the core (abbreviated here as UPL120)was provided with an upper sodium plenum directly above the core to reduce the sodium void reactivity worth. Because of the large fuel pin diameter, average specific fuel power density (31 kW/kg-MOX) of UPL120 is about one half of those of conventional large MOX cores. The reactivity worth of sodium voiding is 6$ in the whole core, and -1$ in the all upper plenums. Initiating phase of ULOF accident in UPL120 under the conditions of nominal design and best estimate analysis resulted in a slightly super-prompt critical power burst. The causes of the super-prompt criticality have been identified twofold: (a) the low specific fuel power density of core reduced the effectiveness of prompt negative reactivity feedback of Doppler and axial fuel expansion effects upon increase in reactor power, and (b) the longer core height compared with conventional 1m cores brought, together with the lower specific power density, a remarkable delay in insertion of negative fuel dispersion reactivity after the onset of fuel disruption in sodium voided subassembly due to the lower linear heat rating in the top portion of the core. During the delay, burst-type fuel failures in sodium un

Calculation of the neutron noise induced by periodic deformations of a large sodium-cooled fast reactor core

International Nuclear Information System (INIS)

Zylbersztejn, F.; Tran, H.N.; Pazsit, I.; Filliatre, P.; Jammes, C.

2014-01-01

The subject of this paper is the calculation of the neutron noise induced by small-amplitude stationary radial variations of the core size (core expansion/compaction, also called core flowering) of a large sodium-cooled fast reactor. The calculations were performed on a realistic model of the European Sodium Fast Reactor (ESFR) core with a thermal output of 3600 MW(thermal), using a multigroup neutron noise simulator. The multigroup cross sections and their fluctuations that represent the core geometry changes for the neutron noise calculations were generated by the code ERANOS. The space and energy dependences of the noise source represented by the core expansion/compaction and the induced neutron noise are calculated and discussed. (authors)

Power flattening and reactivity suppression strategies for the Canadian supercritical water reactor concept

International Nuclear Information System (INIS)

McDonald, M.; Colton, A.; Pencer, J.

2015-01-01

The Canadian supercritical water-cooled reactor (SCWR) is a conceptual heavy water moderated, supercritical light water cooled pressure tube reactor. In contrast to current heavy water power reactors, the Canadian SCWR will be a batch fuelled reactor. Associated with batch fuelling is a large beginning-of-cycle excess reactivity. Furthermore, radial power peaking arising as a consequence of batch refuelling must be mitigated in some way. In this paper, burnable neutron absorber (BNA) added to fuel and absorbing rods inserted into the core are considered for reactivity management and power flattening. A combination of approaches appears adequate to reduce the core radial power peaking, while also providing reactivity suppression. (author)

Evaluation Of Oxide And Silicide Mixed Fuels Of The RSG-GAS Core

International Nuclear Information System (INIS)

Tukiran; Sembiring, Tagor Malem; Suparlina, Lily

2000-01-01

Fuel exchange of the RSG-GAS reactor core from uranium oxide to uranium silicide in the same loading, density, and enrichment, that is 250 gr, 2.98 gr/cm 3 , and 19.75%, respectively, will be performed in-step wise. In every cycle of exchange with 5/1 mode, it is needed to evaluate the parameter of reactor core operation. The parameters of the reactor operation observed are criticality mass of fuels, reactivity balance, and fuel reactivity that give effect to the reactor operation. The evaluation was done at beginning of cycle of the first and second transition core with compared between experiment and calculation results. The experiments were performed at transition core I and II, BOC, and low power. At transition core I, there are 2 silicide fuels (RI-224 and R1-225) in the core and then, added five silicide fuels (R1-226, R1-252, R1-263, and R1-264) to the core, so that there are seven silicide fuels in the transition core II. The evaluation was done based on the experiment of criticality, control rod calibration, fuel reactivity of the RSG-GAS transition core. For inserting 2 silicide fuels in the transition core I dan 7 fuels in the transition core II, the operation of RSG-GAS core fulfilled the safety margin and the parameter of reactor operation change is not occur drastically in experiment and calculation results. So that, the reactor was operated during 36 days at 15 MW, 540 MWD at the first transition core. The general result showed that the parameter of reactor operation change is small so that the fuel exchange from uranium oxide to uranium silicide in the next step can be done

SIMMER analysis of SRI high pressure bubble expansion experiments

International Nuclear Information System (INIS)

Rexroth, P.E.; Suo-Anttila, A.J.

1979-01-01

SIMMER-II was used to analyze the results of the SRI nitrogen bubble expansion experiments. Good agreement was found for all of the experiments analyzed as well as the theoretical isentropic limiting case. Scaling to a full size CRBR reactor reveals no significant scaling effects for the structureless core

Development of Core Design Technology for LMR

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeong Il; Hong, S. G.; Jang, J. W. (and others)

2007-06-15

This report describes the contents of core design technology and computer code system development performed during 2005 and 2006 on the objects of nuclear proliferation resistant core and nuclear fuel basic key technology development security. Also, it is including the future application plans for the results and the developed methodology, important information and the materials acquired in this period. Two core designs with single enrichment were considered for the KALIMER-600 during the first year : 1) the first core uses the non-fuel rods such as B4C, ZrH1.8, and dummy rods, 2) the core using different cladding thickness for each core region (inner, middle, and outer cores) without non-fuel rods to flatten the power distribution. In particular, the latter design was intended to simplify the fuel assembly design by eliminating the heterogeneity. It was found that the proposed design satisfy all of the Gen IV SFR design goals on the cycle length longer than 18 EFPM, fuel discharge burnup larger than 80GWd/t, sodium void worth, conversion ratio, reactivity burnup swing and so on. For this object reactor, the structure integrity outside of reactor is confirmed for the radiation exposure during the plant life according to the result of shielding design and evaluation. The transmutation capability and the core characteristics of sodium cooled fast reactor was also evaluated according to the change of MA amount. The reactivity coefficients for the BN-600 reactor with MA fueled are calculated and the results are compared and evaluated with other participants results. Even though the discrepancies between the results of participants are somewhat large but the K-CORE results are close to the average within a standard deviation. To have the capability of 3-dimensional core dynamic analysis such as analyzing power distribution and reactivity variations according to the asymmetric insertion/withdrawal of control rods, the calculation module for core dynamic parameters was

Comparison of the unitary pole and Adhikari-Sloan expansions in the three nucleon system

International Nuclear Information System (INIS)

Afnan, I.R.; Birrell, N.D.

1977-01-01

The binding energy of 3 H, percentage S-, S'- and D-state probability, and charge form factor of 3 He are calculated using the unitary pole and Adhikari-Sloan separable expansions to the Reid soft core potential. Comparison of the results for the two separable expansions show that the expansion of Adhikari and Sloan has the better convergence property, and the lowest rank expansion considered (equivalent to the unitary pole approximation) gives a good approximation to the binding energy of 3 H and the charge form factor of 3 He, even at large momentum transfer (K 2 -2 ). (Author)

Water flooding criticality study for ZrH flight reactor

International Nuclear Information System (INIS)

Anderson, R.V.

1970-01-01

Five analytical criticality calculations were performed to study the effects of: (1) water reflecting only (no core flooding), (2) water reflection with 10 percent core flooding, (3) water reflection with 35 percent flooding, (4) water reflection plus complete core flooding, and (5) the negative reactivity feedback associated with rapid core expansion induced by a destructive transient. (U.S.)

Core calculational techniques and procedures

International Nuclear Information System (INIS)

Romano, J.J.

1977-10-01

Described are the procedures and techniques employed by B and W in core design analyses of power peaking, control rod worths, and reactivity coefficients. Major emphasis has been placed on current calculational tools and the most frequently performed calculations over the operating power range

Use of compensation assemblies in the first core of SNR-300

International Nuclear Information System (INIS)

Billaux, M.; De Wouters, R.; Pilate, S.; Vandenberg, C.

1975-01-01

For the SNR-300 reactor, the use of thin fuel pins was limited to the first core. A direct consequence of changing from the cycle reloading scheme to a complete irradiation without refueling operation is an increase of the initial excess reactivity and plutonium investment. The new system of special assemblies conceived to compensate for the too high reactivity of the first core is described: fixed absorbers, made of B 4 C pins, and sodium diluents, consisting simply of hollow wrapper tubes [fr

Analysis of the reactivity coefficients of the advanced high-temperature reactor for plutonium and uranium fuels

Energy Technology Data Exchange (ETDEWEB)

Zakova, Jitka [Department of Nuclear and Reactor Physics, Royal Institute of Technology, KTH, Roslagstullsbacken 21, S-10691, Stockholm (Sweden)], E-mail: jitka.zakova@neutron.kth.se; Talamo, Alberto [Nuclear Engineering Division, Argonne National Laboratory, ANL, 9700 South Cass Avenue, Argonne, IL 60439 (United States)], E-mail: alby@anl.gov

2008-05-15

The conceptual design of the advanced high-temperature reactor (AHTR) has recently been proposed by the Oak Ridge National Laboratory, with the intention to provide and alternative energy source for very high temperature applications. In the present study, we focused on the analyses of the reactivity coefficients of the AHTR core fueled with two types of fuel: enriched uranium and plutonium from the reprocessing of light water reactors irradiated fuel. More precisely, we investigated the influence of the outer graphite reflectors on the multiplication factor of the core, the fuel and moderator temperature reactivity coefficients and the void reactivity coefficient for five different molten salts: NaF, BeF{sub 2}, LiF, ZrF{sub 4} and Li{sub 2}BeF{sub 4} eutectic. In order to better illustrate the behavior of the previous parameters for different core configurations, we evaluated the moderating ratio of the molten salts and the absorption rate of the key fuel nuclides, which, of course, are driven by the neutron spectrum. The results show that the fuel and moderator temperature reactivity coefficients are always negative, whereas the void reactivity coefficient can be set negative provided that the fuel to moderator ratio is optimized (the core is undermoderated) and the moderating ratio of the coolant is large.

Analysis of the reactivity coefficients of the advanced high-temperature reactor for plutonium and uranium fuels

International Nuclear Information System (INIS)

Zakova, Jitka; Talamo, Alberto

2008-01-01

The conceptual design of the advanced high-temperature reactor (AHTR) has recently been proposed by the Oak Ridge National Laboratory, with the intention to provide and alternative energy source for very high temperature applications. In the present study, we focused on the analyses of the reactivity coefficients of the AHTR core fueled with two types of fuel: enriched uranium and plutonium from the reprocessing of light water reactors irradiated fuel. More precisely, we investigated the influence of the outer graphite reflectors on the multiplication factor of the core, the fuel and moderator temperature reactivity coefficients and the void reactivity coefficient for five different molten salts: NaF, BeF 2 , LiF, ZrF 4 and Li 2 BeF 4 eutectic. In order to better illustrate the behavior of the previous parameters for different core configurations, we evaluated the moderating ratio of the molten salts and the absorption rate of the key fuel nuclides, which, of course, are driven by the neutron spectrum. The results show that the fuel and moderator temperature reactivity coefficients are always negative, whereas the void reactivity coefficient can be set negative provided that the fuel to moderator ratio is optimized (the core is undermoderated) and the moderating ratio of the coolant is large

Effective neutron temperature measurements in well moderated reactor by the reactivity coefficient method

International Nuclear Information System (INIS)

Raisic, N.; Klinc, T.

1968-11-01

The ratio of the reactivity changes of a nuclear reactor produced by successive introduction of two different neutron absorbers in the reactor core, has been measured and information on effective neutron temperature at a particular point obtained. Boron was used as a l/v absorber and cadmium as an absorber sensiti ve to neutron temperature. Effective neutron temperature distribution has been deduced by moving absorbers across the reactor core and observing the corresponding reactivity changes. (author)

Improvement of JRR-4 core management code system

International Nuclear Information System (INIS)

Izumo, H.; Watanabe, S.; Nagatomi, H.; Hori, N.

2000-01-01

In the modification of JRR-4, the fuel was changed from 93% high enrichment uranium aluminized fuel to 20% low enriched uranium silicide fuel in conformity with the framework of reduced enrichment program on JAERI research reactors. As changing of this, JRR-4 core management code system which estimates excess reactivity of core, fuel burn-up and so on, was improved too. It had been difficult for users to operate the former code system because its input-output form was text-form. But, in the new code system (COMMAS-JRR), users are able to operate the code system without using difficult text-form input. The estimation results of excess reactivity of JRR-4 LEU fuel core were showed very good agreements with the measured value. It is the strong points of this new code system to be operated simply by using the windows form pictures act on a personal workstation equip with the graphical-user-interface (GUI), and to estimate accurately the specific characteristics of the LEU core. (author)

Adaptive Core Simulation Employing Discrete Inverse Theory - Part II: Numerical Experiments

International Nuclear Information System (INIS)

Abdel-Khalik, Hany S.; Turinsky, Paul J.

2005-01-01

Use of adaptive simulation is intended to improve the fidelity and robustness of important core attribute predictions such as core power distribution, thermal margins, and core reactivity. Adaptive simulation utilizes a selected set of past and current reactor measurements of reactor observables, i.e., in-core instrumentation readings, to adapt the simulation in a meaningful way. The companion paper, ''Adaptive Core Simulation Employing Discrete Inverse Theory - Part I: Theory,'' describes in detail the theoretical background of the proposed adaptive techniques. This paper, Part II, demonstrates several computational experiments conducted to assess the fidelity and robustness of the proposed techniques. The intent is to check the ability of the adapted core simulator model to predict future core observables that are not included in the adaption or core observables that are recorded at core conditions that differ from those at which adaption is completed. Also, this paper demonstrates successful utilization of an efficient sensitivity analysis approach to calculate the sensitivity information required to perform the adaption for millions of input core parameters. Finally, this paper illustrates a useful application for adaptive simulation - reducing the inconsistencies between two different core simulator code systems, where the multitudes of input data to one code are adjusted to enhance the agreement between both codes for important core attributes, i.e., core reactivity and power distribution. Also demonstrated is the robustness of such an application

Estimating NIRR-1 burn-up and core life time expectancy using the codes WIMS and CITATION

Science.gov (United States)

Yahaya, B.; Ahmed, Y. A.; Balogun, G. I.; Agbo, S. A.

The Nigeria Research Reactor-1 (NIRR-1) is a low power miniature neutron source reactor (MNSR) located at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria Nigeria. The reactor went critical with initial core excess reactivity of 3.77 mk. The NIRR-1 cold excess reactivity measured at the time of commissioning was determined to be 4.97 mk, which is more than the licensed range of 3.5-4 mk. Hence some cadmium poison worth -1.2 mk was inserted into one of the inner irradiation sites which act as reactivity regulating device in order to reduce the core excess reactivity to 3.77 mk, which is within recommended licensed range of 3.5 mk and 4.0 mk. In this present study, the burn-up calculations of the NIRR-1 fuel and the estimation of the core life time expectancy after 10 years (the reactor core expected cycle) have been conducted using the codes WIMS and CITATION. The burn-up analyses carried out indicated that the excess reactivity of NIRR-1 follows a linear decreasing trend having 216 Effective Full Power Days (EFPD) operations. The reactivity worth of top beryllium shim data plates was calculated to be 19.072 mk. The result of depletion analysis for NIRR-1 core shows that (7.9947 ± 0.0008) g of U-235 was consumed for the period of 12 years of operating time. The production of the build-up of Pu-239 was found to be (0.0347 ± 0.0043) g. The core life time estimated in this research was found to be 30.33 years. This is in good agreement with the literature

Analysis of Doppler effect measurement in FCA cores using JENDL-3.2 library

International Nuclear Information System (INIS)

Okajima, Shigeaki

1996-01-01

For the evaluation of the calculation accuracy of the 238 U Doppler effect using JENDL-3.2 library, the previously measured Doppler reactivity worths in the FCA were systematically analyzed. In the analysis the Doppler reactivity worth was calculated by a first order perturbation theory. The calculated results were compared with those using JENDL-3.1 library. The JENDL-3.2 calculation in MOX fuel mock-up cores agrees well with the experimental values within the experimental error. In U-235/Pu fuel cores, the JENDL-3.2 calculation gives 12-15% larger Doppler reactivity worths than the JENDL-3.1 calculation. (author)

An amino-terminal segment of hantavirus nucleocapsid protein presented on hepatitis B virus core particles induces a strong and highly cross-reactive antibody response in mice

International Nuclear Information System (INIS)

Geldmacher, Astrid; Skrastina, Dace; Petrovskis, Ivars; Borisova, Galina; Berriman, John A.; Roseman, Alan M.; Crowther, R. Anthony; Fischer, Jan; Musema, Shamil; Gelderblom, Hans R.; Lundkvist, Aake; Renhofa, Regina; Ose, Velta; Krueger, Detlev H.; Pumpens, Paul; Ulrich, Rainer

2004-01-01

Previously, we have demonstrated that hepatitis B virus (HBV) core particles tolerate the insertion of the amino-terminal 120 amino acids (aa) of the Puumala hantavirus nucleocapsid (N) protein. Here, we demonstrate that the insertion of 120 amino-terminal aa of N proteins from highly virulent Dobrava and Hantaan hantaviruses allows the formation of chimeric core particles. These particles expose the inserted foreign protein segments, at least in part, on their surface. Analysis by electron cryomicroscopy of chimeric particles harbouring the Puumala virus (PUUV) N segment revealed 90% T = 3 and 10% T = 4 shells. A map computed from T = 3 shells shows additional density splaying out from the tips of the spikes producing the effect of an extra shell of density at an outer radius compared with wild-type shells. The inserted Puumala virus N protein segment is flexibly linked to the core spikes and only partially icosahedrally ordered. Immunisation of mice of two different haplotypes (BALB/c and C57BL/6) with chimeric core particles induces a high-titered and highly cross-reactive N-specific antibody response in both mice strains

RELAP5-3D code validation of RBMK-1500 reactor reactivity measurement transients

International Nuclear Information System (INIS)

Kaliatka, Algirdas; Bubelis, Evaldas; Uspuras, Eugenijus

2003-01-01

This paper deals with the modeling of transients taking place during the measurements of the void and fast power reactivity coefficients performed at Ignalina NPP. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Following the simulation of the two above mentioned transients with RELAP5-3D code, a conclusion was made that the obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data. Despite of the small differences, RELAP5-3D code predicts reactivity and the total reactor core power behavior during the transients in a reasonable manner. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core

Neutron dynamics of fast-spectrum dedicated cores for waste transmutation

International Nuclear Information System (INIS)

Massara, S.

2002-04-01

Among different scenarios achieving minor actinide transmutation, the possibility of double strata scenarios with critical, fast spectrum, dedicated cores must be checked and quantified. In these cores, the waste fraction has to be at the highest level compatible with safety requirements during normal operation and transient conditions. As reactivity coefficients are poor in such critical cores (low delayed neutron fraction and Doppler feed-back, high coolant void coefficient), their dynamic behaviour during transient conditions must be carefully analysed. Three nitride-fuel configurations have been analysed: two liquid metal-cooled (sodium and lead) and a particle-fuel helium-cooled one. A dynamic code, MAT4 DYN, has been developed during the PhD thesis, allowing the study of loss of flow, reactivity insertion and loss of coolant accidents, and taking into account two fuel geometries (cylindrical and spherical) and two thermal-hydraulics models for the coolant (incompressible for liquid metals and compressible for helium). Dynamics calculations have shown that if the fuel nature is appropriately chosen (letting a sufficient margin during transients), this can counterbalance the bad state of reactivity coefficients for liquid metal-cooled cores, thus proving the interest of this kind of concept. On the other side, the gas-cooled core dynamics is very badly affected by the high value of the helium void coefficient (which is a consequence of the choice of a hard spectrum), this effect being amplified by the very low thermal inertia of particle-fuel design. So, a new kind of concept should be considered for a helium-cooled fast-spectrum dedicated core. (authors)

Monitoring of the temperature reactivity coefficient at the PWR nuclear plant

International Nuclear Information System (INIS)

Kostic, Lj.

1996-01-01

For monitoring temperature coefficient of reactivity of pressurized water reactor a method based on the correction of fluctuation in signals of i-core neutron detectors and core-exit thermocouples and neural network paradigm is used it is shown that the moderator temperature coefficient of relativity can be predicted with the aid of the back propagation neural network technique by measuring the frequency response function between the in-core neutron flux and the core-exit coolant temperature

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)

Isothermal temperature reactivity coefficient measurement in TRIGA reactor

International Nuclear Information System (INIS)

Zagar, T.; Ravnik, M.; Trkov, A.

2002-01-01

Direct measurement of an isothermal temperature reactivity coefficient at room temperatures in TRIGA Mark II research reactor at Jozef Stefan Institute in Ljubljana is presented. Temperature reactivity coefficient was measured in the temperature range between 15 o C and 25 o C. All reactivity measurements were performed at almost zero reactor power to reduce or completely eliminate nuclear heating. Slow and steady temperature decrease was controlled using the reactor tank cooling system. In this way the temperatures of fuel, of moderator and of coolant were kept in equilibrium throughout the measurements. It was found out that TRIGA reactor core loaded with standard fuel elements with stainless steel cladding has small positive isothermal temperature reactivity coefficient in this temperature range.(author)

Some concept for the TRIGA core design

International Nuclear Information System (INIS)

Aizawa, Otohiko

1994-01-01

There is the research reactor called TRIGA Mark-2 of 100 kW in Atomic Energy Research Laboratory, Musashi Institute of Technology. Recently, while the various calculations on the core were carried out, the author became aware of that this TRIGA core was designed at that time with excellent consideration. The reason for that is, although fuel is arranged in simple concentric circular state at a glance, it was known that in reality, this is the modification of the hexagonal core of triangular lattice. In the examination of square lattice fuel arrangement, the reactivity was calculated by using the gap between fuel rods as the parameter and by using ENDF/B-4 library and Monte Carlo code Keno-5. It is known that the design of the lattice with maximum reactivity cannot be done by the square lattice. The similar examination was carried out on triangular lattice, and it was found that the gap between fuel rods of 4 mm is the optimal design. The average neutron energy spectra in the fuel rods of the TRIGA Mark-2 core agreed considerably well with the energy spectra at 4.16 cm fuel rod pitch in triangular hexagonal core. In the reactor of about 100 kW, even if the gap between fuel rods is less than 4 mm, heat removal is sufficiently possible. (K.I.)

Enhanced thermal expansion control rod drive lines for improving passive safety of fast reactors

International Nuclear Information System (INIS)

Edelmann, M.; Baumann, W.; Kuechle, M.; Kussmaul, G.; Vaeth, W.; Bertram, A.

1992-01-01

The paper presents a device for increasing the thermal expansion effect of control rod drive lines on negative reactivity feedback in fast reactors. The enhanced thermal expansion of this device can be utilized for both passive rod drop and forced insertion of absorbers in unprotected transients, e.g. ULOF. In this way the reactor is automatically brought into a permanently subcritical state and temperatures are kept well below the boiling point of the coolant. A prototype of such a device called ATHENa (German: Shut-down by THermal Expansion of Na) is presently under construction and will be tested. The paper presents the principle, design features and thermal properties of ATHENs as well as results of reactor dynamics calculations of ULOF's for EFR with enhanced thermal expansion control rod drive lines. (author)

Conceptual core model for the reactor core test

International Nuclear Information System (INIS)

Swenson, L.D.

1970-01-01

Several design options for the ZrH Flight System Reactor were investigated which involved tradeoffs of core excess reactivity, reactor control, coolant mixing and cladding thickness. A design point was selected which is to be the basis for more detailed evaluation in the preliminary design phase. The selected design utilizes 295 elements with 0.670 inch element-to-element pitch, 32 mil thick Incoloy cladding, 18.00 inches long fuel meat, hydrogen content of 6.3 x 10 22 atoms/cc fuel, 10.5 w/o uranium, and a spiraled fin configuration with alternate elements having fins with spiral to the right, spiral to the left, and no fin at all (R-L-N fin configuration). Fin height is 30 mils for the center region of the core and 15 mils for the outer region. (U.S.)

Neutron spectrum effects on TRU recycling in Pb-Bi cooled fast reactor core

International Nuclear Information System (INIS)

Kim, Yong Nam; Kim, Jong Kyung; Park, Won Seok

2003-01-01

This study is intended to evaluate the dependency of TRU recycling characteristics on the neutron spectrum shift in a Pb-Bi cooled core. Considering two Pb-Bi cooled cores with the soft and the hard spectrum, respectively, various characteristics of the recycled core are carefully examined and compared with each other. Assuming very simplified fuel cycle management with the homogeneous and single region fuel loading, the burnup calculations are performed until the recycled core reached to the (quasi-) equilibrium state. The mechanism of TRU recycling toward the equilibrium is analyzed in terms of burnup reactivity and the isotopic compositions of TRU fuel. In the comparative analyses, the difference in the recycling behavior between the two cores is clarified. In addition, the basic safety characteristics of the recycled core are also discussed in terms of the Doppler coefficient, the coolant loss reactivity coefficient, and the effective delayed neutron fraction

(Electronic structure and reactivities of transition metal clusters)

Energy Technology Data Exchange (ETDEWEB)

1992-01-01

The following are reported: theoretical calculations (configuration interaction, relativistic effective core potentials, polyatomics, CASSCF); proposed theoretical studies (clusters of Cu, Ag, Au, Ni, Pt, Pd, Rh, Ir, Os, Ru; transition metal cluster ions; transition metal carbide clusters; bimetallic mixed transition metal clusters); reactivity studies on transition metal clusters (reactivity with H{sub 2}, C{sub 2}H{sub 4}, hydrocarbons; NO and CO chemisorption on surfaces). Computer facilities and codes to be used, are described. 192 refs, 13 figs.

Reactor Core Design and Analysis for a Micronuclear Power Source

Directory of Open Access Journals (Sweden)

Hao Sun

2018-03-01

Full Text Available Underwater vehicle is designed to ensure the security of country sea boundary, providing harsh requirements for its power system design. Conventional power sources, such as battery and Stirling engine, are featured with low power and short lifetime. Micronuclear reactor power source featured with higher power density and longer lifetime would strongly meet the demands of unmanned underwater vehicle power system. In this paper, a 2.4 MWt lithium heat pipe cooled reactor core is designed for micronuclear power source, which can be applied for underwater vehicles. The core features with small volume, high power density, long lifetime, and low noise level. Uranium nitride fuel with 70% enrichment and lithium heat pipes are adopted in the core. The reactivity is controlled by six control drums with B4C neutron absorber. Monte Carlo code MCNP is used for calculating the power distribution, characteristics of reactivity feedback, and core criticality safety. A code MCORE coupling MCNP and ORIGEN is used to analyze the burnup characteristics of the designed core. The results show that the core life is 14 years, and the core parameters satisfy the safety requirements. This work provides reference to the design and application of the micronuclear power source.

Dependence of calculated void reactivity on film-boiling representation

International Nuclear Information System (INIS)

Whitlock, J.; Garland, W.

1992-01-01

Partial voiding of a fuel channel can lead to complicated neutronic analysis, because of highly nonuniform spatial distributions. An investigation of the distribution dependence of void reactivity in a Canada deuterium uranium (CANDU) lattice, specifically in the regime of film boiling, was done. Although the core is not expected to be critical at the time of sheath dryout, this study augments current knowledge of void reactivity in this type of lattice

A feasibility study on wavelet transform for reactivity coefficient estimation

International Nuclear Information System (INIS)

Shimazu, Yoichiro

2000-01-01

Recently, a new method using Fourier transform has been introduced in place of the conventional method in order to reduce the time required for the measurement of moderator temperature coefficient in domestic PWRs. The basic concept of these methods is to eliminate noise in the reactivity signal. From this point of view, wavelet analysis is also known as an effective method. In this paper, we tried to apply this method to estimate reactivity coefficients of a nuclear reactor. The basic idea of the reactivity coefficient estimation is to analyze the ratios themselves of the corresponding expansion coefficients of the wavelet transform of the signals of reactivity and the relevant parameter. The concept requires no inverse wavelet transform. Based on numerical simulations, it is found that the method can reasonably estimate reactivity coefficient, for example moderator temperature coefficient, with less length of time sequence data than those required for Fourier transform method. We will continue this study to examine the validity of the estimation procedure for the actual reactor data and further to estimate the other reactivity coefficients. (author)

The causal perturbation expansion revisited: Rescaling the interacting Dirac sea

International Nuclear Information System (INIS)

Finster, Felix; Grotz, Andreas

2010-01-01

The causal perturbation expansion defines the Dirac sea in the presence of a time-dependent external field. It yields an operator whose image generalizes the vacuum solutions of negative energy and thus gives a canonical splitting of the solution space into two subspaces. After giving a self-contained introduction to the ideas and techniques, we show that this operator is, in general, not idempotent. We modify the standard construction by a rescaling procedure giving a projector on the generalized negative-energy subspace. The resulting rescaled causal perturbation expansion uniquely defines the fermionic projector in terms of a series of distributional solutions of the Dirac equation. The technical core of the paper is to work out the combinatorics of the expansion in detail. It is also shown that the fermionic projector with interaction can be obtained from the free projector by a unitary transformation. We finally analyze the consequences of the rescaling procedure on the light-cone expansion.

The causal perturbation expansion revisited: Rescaling the interacting Dirac sea

Science.gov (United States)

Finster, Felix; Grotz, Andreas

2010-07-01

The causal perturbation expansion defines the Dirac sea in the presence of a time-dependent external field. It yields an operator whose image generalizes the vacuum solutions of negative energy and thus gives a canonical splitting of the solution space into two subspaces. After giving a self-contained introduction to the ideas and techniques, we show that this operator is, in general, not idempotent. We modify the standard construction by a rescaling procedure giving a projector on the generalized negative-energy subspace. The resulting rescaled causal perturbation expansion uniquely defines the fermionic projector in terms of a series of distributional solutions of the Dirac equation. The technical core of the paper is to work out the combinatorics of the expansion in detail. It is also shown that the fermionic projector with interaction can be obtained from the free projector by a unitary transformation. We finally analyze the consequences of the rescaling procedure on the light-cone expansion.

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)

Alteration of helical vortex core without change in flow topology

DEFF Research Database (Denmark)

Velte, Clara Marika; Okulov, Valery; Hansen, Martin Otto Laver

2011-01-01

topology. The helical symmetry as such is preserved, although the characteristic parameters of helical symmetry of the vortex core transfer from a smooth linear variation to a different trend under the influence of a non-uniform pressure gradient, causing an increase in helical pitch without changing its......The abrupt expansion of the slender vortex core with changes in flow topology is commonly known as vortex breakdown. We present new experimental observations of an alteration of the helical vortex core in wall bounded turbulent flow with abrupt growth in core size, but without change in flow...

Critical experiments of JMTRC MEU cores

International Nuclear Information System (INIS)

Nagaoka, Y.; Takeda, K.; Shimakawa, S.; Koike, S.; Oyamada, R.

1984-01-01

The JMTRC, the critical facility of the Japan Materials Testing Reactor (JMTR), went critical on August 29, 1983, with 14 medium enriched uranium (MEU, 45%) fuel elements. Experiments are now being carried out to measure the change in various reactor characteristics between the previous HEU core and the new MEU fueled core. This paper describes the results obtained thus far on critical mass, excess reactivity, control rod worths and flux distribution, including preliminary neutronics calculations for the experiments using the SRAC code. (author)

Space Flight-Induced Reactivation of Latent Epstein-Barr Virus

Science.gov (United States)

Stowe, Raymond P.; Barrett, Alan D. T.; Pierson, Duane L.

2001-01-01

Reactivation of latent Epstein-Barr virus (EBV) may be an important threat to crew health during extended space missions. Decreased cellular immune function has been reported both during and after space flight. Preliminary studies have demonstrated increased EBV shedding in saliva as well as increased antibody titers to EBV lytic proteins. We hypothesize that the combined effects of microgravity along with associated physical and psychological stress will decrease EBV-specific T-cell immunity and reactivate latent EBV in infected B-lymphocytes. If increased virus production and clonal expansion of infected B-lymphocytes are detected, then pharmacological measures can be developed and instituted prior to onset of overt clinical disease. More importantly, we will begin to understand the basic mechanisms involved in stress-induced reactivation of EBV in circulating B-lymphocytes.

Expansion of mycobacterium-reactive gamma delta T cells by a subset of memory helper T cells.

Science.gov (United States)

Vila, L M; Haftel, H M; Park, H S; Lin, M S; Romzek, N C; Hanash, S M; Holoshitz, J

1995-04-01

Human gamma delta T cells expressing the V gamma 9/V delta 2 T-cell receptor have been previously found to proliferate in response to certain microorganisms and to expand throughout life, presumably because of extrathymic activation by foreign antigens. In vitro expansion of V gamma 9/V delta 2 cells by mycobacteria has been previously shown to be dependent on accessory cells. In order to gain an insight into the mechanisms involved in the expansion of these cells, we have undertaken to identify the peripheral blood subset of cells on which proliferation of V gamma 9/V delta 2 cells in response to mycobacteria is dependent. Contrary to their role in antigen presentation to alpha beta T cells, professional antigen-presenting cells, such as monocytes, B cells, and dendritic cells, were unable to provide the cellular support for the expansion of V gamma 9/V delta 2 cells. Selective depletion of T-cell subsets, as well as the use of highly purified T-cell populations, indicated that the only subset of peripheral blood cells that could expand V gamma 9/V delta 2 cells were CD4+ CD45RO+ CD7- alpha beta T cells. These cells underwent distinct intracellular signaling events after stimulation with the mycobacterial antigen. Expansion of V gamma 9/V delta 2 cells by alpha beta T cells was dependent on cell-cell contact. This is the first evidence that a small subset of the memory helper T-cell population is exclusively responsible for the peripheral expansion of V gamma 9/V delta 2 cells. These data illustrate a unique aspect of antigen recognition by gamma delta T cells and provide new means to study their immune defense role.

A detailed neutronics comparison of the university of Florida training reactor (UFTR) current HEU and proposed LEU cores

International Nuclear Information System (INIS)

Dionne, B.; Haghighat, A.; Yi, C.; Smith, R.; Ghita, G.; Manalo, K.; Sjoden, G.; Huh, J.; Baciak, J.; Mock, T.; Wenner, M.; Matos, J.; Stillman, J.

2006-01-01

For over 35 years, the UFTR highly-enriched core has been safely operated. As part of the Reduced Enrichment for Research and Test Reactors Program, the core is currently being converted to low-enriched uranium fuel. The analyses presented in this paper were performed to verify that, from a neutronic perspective, a proposed low-enriched core can be operated as safely and as effectively as the highly-enriched core. Detailed Monte Carlo criticality calculations are performed to determine: i) Excess reactivity for different core configurations, ii) Individual integral blade worth and shutdown margin, iii) Reactivity coefficients and kinetic parameters, and iv) Flux profiles and core six-factor formula parameters. (authors)

Plutonium cores of zenith

Energy Technology Data Exchange (ETDEWEB)

Barclay, F R; Cameron, I R; Drageset, A; Freemantle, R G; Wilson, D J

1965-03-15

The report describes a series of experiments carried out with plutonium fuel in the heated zero power reactor ZENITH, with the aim of testing current theoretical methods, with particular reference to excess reactivity, temperature coefficients, differential spectrum and reaction rate distributions. Two cores of widely different fissile/moderator atom ratios were loaded in order to test the theory under significantly varied spectrum conditions.

Calculation-measurement comparison for control rods reactivity in RA-3 nuclear reactor

International Nuclear Information System (INIS)

Estryk, Guillermo; Gomez, Angel

2002-01-01

The RA-3 Nuclear Reactor of the Atomic Energy National Commission from Argentina, begun working with high enrichment fuel elements in 1967, and turned to low enrichment by 1990. During 1999 it was found out that several fuel elements had problems, so more than 50 % of them had to be removed from the core. Because of this, it was planned to go from core 93 to core 94 with special care from nuclear safety point of view. Core 94 was preceded by other five, T-1 to T-5, only as transitory ones. The care implied several nuclear parameters measurements: core reactivity excess, calibration of control rods, etc. Calculations were performed afterwards to simulate those measurements using the neutron diffusion code PUMA. The comparison shows a good agreement for more than 80% of the cases with differences lower than 10% in reactivity. The greatest differences were found in the last part of the control rods calibration and a better calculation of cell constants is planned to be done in order to improve the adjustment. (author)

Inhibitory Effect of Waste Glass Powder on ASR Expansion Induced by Waste Glass Aggregate

Directory of Open Access Journals (Sweden)

Shuhua Liu

2015-10-01

Full Text Available Detailed research is carried out to ascertain the inhibitory effect of waste glass powder (WGP on alkali-silica reaction (ASR expansion induced by waste glass aggregate in this paper. The alkali reactivity of waste glass aggregate is examined by two methods in accordance with the China Test Code SL352-2006. The potential of WGP to control the ASR expansion is determined in terms of mean diameter, specific surface area, content of WGP and curing temperature. Two mathematical models are developed to estimate the inhibitory efficiency of WGP. These studies show that there is ASR risk with an ASR expansion rate over 0.2% when the sand contains more than 30% glass aggregate. However, WGP can effectively control the ASR expansion and inhibit the expansion rate induced by the glass aggregate to be under 0.1%. The two mathematical models have good simulation results, which can be used to evaluate the inhibitory effect of WGP on ASR risk.

Development of a Two-dimensional Thermohydraulic Hot Pool Model and ITS Effects on Reactivity Feedback during a UTOP in Liquid Metal Reactors

International Nuclear Information System (INIS)

Lee, Yong Bum; Jeong, Hae Yong; Cho, Chung Ho; Kwon, Young Min; Ha, Kwi Seok; Chang, Won Pyo; Suk, Soo Dong; Hahn, Do Hee

2009-01-01

The existence of a large sodium pool in the KALIMER, a pool-type LMR developed by the Korea Atomic Energy Research Institute, plays an important role in reactor safety and operability because it determines the grace time for operators to cope with an abnormal event and to terminate a transient before reactor enters into an accident condition. A two-dimensional hot pool model has been developed and implemented in the SSC-K code, and has been successfully applied for the assessment of safety issues in the conceptual design of KALIMER and for the analysis of anticipated system transients. The other important models of the SSC-K code include a three-dimensional core thermal-hydraulic model, a reactivity model, a passive decay heat removal system model, and an intermediate heat transport system and steam generation system model. The capability of the developed two-dimensional hot pool model was evaluated with a comparison of the temperature distribution calculated with the CFX code. The predicted hot pool coolant temperature distributions obtained with the two-dimensional hot pool model agreed well with those predicted with the CFX code. Variations in the temperature distribution of the hot pool affect the reactivity feedback due to an expansion of the control rod drive line (CRDL) immersed in the pool. The existing CRDL reactivity model of the SSC-K code has been modified based on the detailed hot pool temperature distribution obtained with the two-dimensional pool model. An analysis of an unprotected transient over power with the modified reactivity model showed an improved negative reactivity feedback effect

Preliminary concept of a zero power nuclear reactor core

International Nuclear Information System (INIS)

Mai, Luiz Antonio; Siqueira, Paulo de Tarso D.

2011-01-01

The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

Preliminary concept of a zero power nuclear reactor core

Energy Technology Data Exchange (ETDEWEB)

Mai, Luiz Antonio; Siqueira, Paulo de Tarso D., E-mail: lamai@ipen.b, E-mail: ptsiquei@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

Long term storage effects of irradiated fuel elements on power distribution and reactivity

Energy Technology Data Exchange (ETDEWEB)

Ponzoni Filho, P.; Sato, Sadakatu; Santos, Teresinha Ipojuca Cardoso T.I.C.; Fernandes Vanderlei Borba [FURNAS, Rio de Janeiro, RJ (Brazil); Fetterman, R.J. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

1995-12-31

The ALPHA/PHOENIX-P/ANC (APA) code package was used to calculate the pin by pin power distribution and reactivity for Angra 1 Power Plant, Cycle 5. The Angra 1 Cycle 5 core was loaded with several irradiated fuel elements which were stored in the Spent Fuel Pool (SFP) for more than 8 years. Generally, neutronic codes take into account the buildup and depletion of just a few key fission, products such as Sm-149. In this paper it is shown that the buildup effects of other fission products must be considered for fuel which has been out of the core for significant periods of time. Impacts of these other fission products can change core reactivity and power distribution. (author). 3 refs, 4 figs, 4 tabs.

Long term storage effects of irradiated fuel elements on power distribution and reactivity

International Nuclear Information System (INIS)

Ponzoni Filho, P.; Sato, Sadakatu; Santos, Teresinha Ipojuca Cardoso T.I.C.; Fernandes Vanderlei Borba; Fetterman, R.J.

1995-01-01

The ALPHA/PHOENIX-P/ANC (APA) code package was used to calculate the pin by pin power distribution and reactivity for Angra 1 Power Plant, Cycle 5. The Angra 1 Cycle 5 core was loaded with several irradiated fuel elements which were stored in the Spent Fuel Pool (SFP) for more than 8 years. Generally, neutronic codes take into account the buildup and depletion of just a few key fission, products such as Sm-149. In this paper it is shown that the buildup effects of other fission products must be considered for fuel which has been out of the core for significant periods of time. Impacts of these other fission products can change core reactivity and power distribution. (author). 3 refs, 4 figs, 4 tabs

Transient bowing of core assemblies in advanced liquid metal fast reactors

International Nuclear Information System (INIS)

Kamal, S.A.; Orechwa, Y.

1986-01-01

Two alternative core restraint concepts are considered for a conceptual design of a 900 MWth liquid metal fast reactor core with a heterogeneous layout. The two concepts, known as limited free bowing and free flowering, are evaluated based on core bowing criteria that emphasize the enhancement of inherent reactor safety. The core reactivity change during a postulated loss of flow transient is calculated in terms of the lateral displacements and displacement-reactivity-worths of the individual assemblies. The NUBOW-3D computer code is utilized to determine the assembly deformations and interassembly forces that arise when the assemblies are subjected to temperature gradients and irradiation induced creep and swelling during the reactor operation. The assembly ducts are made of the ferritic steel HT-9 and remain in the reactor core for four-years at full power condition. Whereas both restraint systems meet the bowing criteria, a properly designed limited free bowing system appears to be more advantageous than a free flowering system from the point of view of enhancing the reactor inherent safety

Transmission expansion cost allocation based on cooperative game theory for congestion relief

Energy Technology Data Exchange (ETDEWEB)

Erli, Ge; Takahasi, Kazuhiro; Kurihara, Ikuo [Central Research Inst. of Electric Power Industry, Tokyo (Japan); Luonan Chen [Osaka Sangyo Univ., Osaka (Japan)

2005-01-01

In conventional power systems, upstream and downstream of power were distinct. However, due to the competition, power injection and sink can appear at unexpected locations, and cost sharing for such a new power system configuration must be considered. This paper proposes a scheme for transmission expansion cost allocation among electric market participants by using Core and Nucleolus concepts of game theory, which are developed particularly for the transmission users. A solution of the n-person cooperative game is adopted to distribute the line transmission expansion cost among the players. Congestion is assumed to be the transmission constraint, and expansion of transmission line is expected to relieve transmission congestion. A case study is illustrated to demonstrate the proposed method. (Author)

Digital instrument for reactivity measurements in a nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Chwaszczewski, S [Institute of Nuclear Research, Warsaw (Poland)

1979-07-01

An instrument for digital determination of the reactivity in nuclear reactors is described. It is based on the CAMAC standard apparatus, suitable for the use of pulse or current type neutron detectors and operates with prompt response and an output signal proportional to the core neutron flux. The measured data of neutron flux and reactivity can be registered by a digital display unit, an indicator, or, by request of the operator, a paper type punch. The algorithms used for reactivity calculation are considered and the results of numerical studies on those algorithms are discussed. The instrument has been used for determining the reactivity of the control elements in the fast-thermal assembly ANNA and in the research reactor MARIA. Some results of these measurements are given.

SCRAM reactivity calculations with the KIKO3D code

International Nuclear Information System (INIS)

Hordosy, G.; Kerszturi, A.; Maraczy, Cs.; Temesvari, E.

1999-01-01

Discrepancies between calculated static reactivities and measured reactivities evaluated with reactivity meters led to investigating SCRAM with the KIKO3D dynamic code, The time and space dependent neutron flux in the reactor core during the rod drop measurement was calculated by the KIKO3D nodal diffusion code. For calculating the ionisation chamber signals the Green function technique was applied. The Green functions of ionisation chambers were evaluated via solving the neutron transport equation in the reflector regions with the MCNP Monte Carlo code. The detector signals during asymmetric SCRAM measurements were calculated and compared with measured data using the inverse point kinetics transformation. The sufficient agreement validates the KIKO3D code to determine the reactivities after SCRAM. (Authors)

Simple kinetic theory model of reactive collisions. IV. Laboratory fixed orientational cross sections

International Nuclear Information System (INIS)

Evans, G.T.

1987-01-01

The differential orientational cross section, obtainable from molecular beam experiments on aligned molecules, is calculated using the line-of-normals model for reactive collisions involving hard convex bodies. By means of kinetic theory methods, the dependence of the cross section on the angle of attack γ 0 is expressed in a Legendre function expansion. Each of the Legendre expansion coefficients is given by an integral over the molecule-fixed cross section and functions of the orientation dependent threshold energy

The CRDL model of SSC-K code for the safety improvement of a pool-type liquid metal-cooled reactor

International Nuclear Information System (INIS)

Jung, H. Y.; Ha, K. S.; Jang, W. P.; Hu, S.; Lee, Y. B.

2004-01-01

With the increased thermal power of KALIMER-600, it becomes important to model accurately the reactivity feedback effects due to the thermal expansion of a fuel rod and internal structure during a transient. In KALIMER design, the fuel axial expansion, core radial expansion, and the control rod drive line/reactor vessel (CRDL/RV) thermal expansion are the important reactivity feedback mechanisms. It is required to develop a more detailed CRDL/RV model for the accurate analysis of the KALIMER-600 transient because the control rod drive line of 9.5 m is immersed in the hot pool. For this a new CRDL/RV model was developed to model the effect of expansion of CRDL utilizing the temperature distribution obtained with the hot-pool 2-D model of SSC-K code. It is estimated that the developed model describes more realistically the negative reactivity insertion effect due to the initial temperature change during the UTOP transient of KALIMER-150

JOYO MK-III performance test. Criticality test, excess reactivity measurement and burn-up coefficient measurement

International Nuclear Information System (INIS)

Maeda, Shigetaka; Sekine, Takashi; Kitano, Akihiro; Nagasaki, Hideaki

2005-03-01

The MK-III performance test began in June 2003 to fully characterize the upgraded core and heat transfer system of the experimental fast reactor JOYO. This paper describes the results of the approach to criticality, the excess reactivity evaluation and the burn-up coefficient measurement. In the approach to criticality test, the MK-III core achieved initial criticality at the control rod bank position of 412.8 mm on 14:03 July 2nd, 2003. Because the replacement of the outer two rows of reflector subassemblies with shielding subassemblies reduced the source range monitor signals by a factor of 3 at the same reactor power compared with those in the MK-II core, we measured the change of the monitor's response and determined the count rate 2x10 4 cps.' as an appropriate value judging the zero power criticality. In the excess reactivity evaluation, the zero power excess reactivity at 250degC was 2.99±0.10%Δk/kk' based on the measured critical rod bank position and the measured control rod worths. The predicted value by the JOYO core management code system HESTIA was 3.13±0.16%Δk/kk', showing good agreement with the measured value. The measured excess reactivity was within the safety requirement limit. In the burn-up coefficient measurement, the excess reactivity change versus the reactor burn-up was evaluated. The measurement method adopted was to measure the control rod positions during the rated power operation. A value of -2.12x10 -4 Δk/kk'/MWd was obtained as a measured burn-up coefficient. The value calculated by HESTIA was -2.12x10 -4 Δk/kk'/MWd, and it agreed well with the measured value. All technical safety requirements for MK-III core were satisfied and the calculation accuracy of the core management code system HESTIA was confirmed. (author)

Neutronic calculations of PARR-1 cores using LEU silicide fuel

International Nuclear Information System (INIS)

Arshad, M.; Bakhtyar, S.; Hayat, T.; Salahuddin, A.

1991-08-01

Detailed neutronic calculations have been carried out for different PARR-1 cores utilizing low enriched uranium (LEU) silicide fuel and operating at an upgraded power of 9 MW. The calculations include the search for critical loadings in open and stall ends of the pool, neutronic analysis of the first full equilibrium core and calculations cores. The burnup study of inventory have also been carried out. Further, the reactivity coefficients of the first full power operation core are evaluated for use in the accident analysis. 14 figs. (author)

Experimental estimation of moderator temperature coefficient of reactivity of the IPEN/MB-01 research reactor

International Nuclear Information System (INIS)

Silva, Rubens C. da; Bitelli, Ulysses D.; Mura, Luiz Ernesto C.

2017-01-01

The aim of this article is to present the procedure for the experimental estimation of the Moderator Temperature Coefficient of Reactivity of the IPEN/MB-01 Research Reactor, a parameter that has an important role in the physics and the control operations of any reactor facility. At the experiment, the IPEN/MB-01 reactor went critical at the power of 1W (1% of its total power), and whose core configuration was 28 x 26 rectangular array of UO_2 fuel rods, inside a light water (moderator) tank. In addition, there was a heavy water (D_2O) reflector installed in the West side of the core to obtain an adequate neutron reflection along the experiment. The moderator temperature was increased in steps of 4 °C, and the measurement of the mean moderator temperature was acquired using twelve calibrated thermocouples, placed around the reactor core. As a result, the mean value of -4.81 pcm/°C was obtained for such coefficient. The curves of ρ(T) (Reactivity x Temperature) and α"M_T(T)(Moderator Temperature Coefficient of Reactivity x Temperature) were developed using data from an experimental measurement of the integral reactivity curves through the Stable Period and Inverse Kinetics Methods, that was carried out at the reactor with the same core configuration. Such curves were compared and showed a very similar behavior between them. (author)

Intrinsically secure fast reactors with dense cores

International Nuclear Information System (INIS)

Slessarev, Igor

2007-01-01

Secure safety, resistance to weapons material proliferation and problems of long-lived wastes remain the most important 'painful points' of nuclear power. Many innovative reactor concepts have been developed aimed at a radical enhancement of safety. The promising potential of innovative nuclear reactors allows for shifting accents in current reactor safety 'strategy' to reveal this worth. Such strategy is elaborated focusing on the priority for intrinsically secure safety features as well as on sure protection being provided by the first barrier of defence. Concerning the potential of fast reactors (i.e. sodium cooled, lead-cooled, etc.), there are no doubts that they are able to possess many favourable intrinsically secure safety features and to lay the proper foundation for a new reactor generation. However, some of their neutronic characteristics have to be radically improved. Among intrinsically secure safety properties, the following core parameters are significantly important: reactivity margin values, reactivity feed-back and coolant void effects. Ways of designing intrinsically secure safety features in fast reactors (titled hereafter as Intrinsically Secure Fast Reactors - ISFR) can be found in the frame of current reactor technologies by radical enhancement of core neutron economy and by optimization of core compositions. Simultaneously, respecting resistance to proliferation, by using non-enriched fuel feed as well as a core breeding gain close to zero, are considered as the important features (long-lived waste problems will be considered in a separate paper). This implies using the following reactor design options as well as closed fuel cycles with natural U as the reactor feed: ·Ultra-plate 'dense cores' of the ordinary (monolithic) type with negative total coolant void effects. ·Modular type cores. Multiple dense modules can be embedded in the common reflector for achieving the desired NPP total power. The modules can be used also independently (as

Advanced computational methods for the assessment of reactor core behaviour during reactivity initiated accidents. Final report; Fortschrittliche Rechenmethoden zum Kernverhalten bei Reaktivitaetsstoerfaellen. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Pautz, A.; Perin, Y.; Pasichnyk, I.; Velkov, K.; Zwermann, W.; Seubert, A.; Klein, M.; Gallner, L.; Krzycacz-Hausmann, B.

2012-05-15

The document at hand serves as the final report for the reactor safety research project RS1183 ''Advanced Computational Methods for the Assessment of Reactor Core Behavior During Reactivity-Initiated Accidents''. The work performed in the framework of this project was dedicated to the development, validation and application of advanced computational methods for the simulation of transients and accidents of nuclear installations. These simulation tools describe in particular the behavior of the reactor core (with respect to neutronics, thermal-hydraulics and thermal mechanics) at a very high level of detail. The overall goal of this project was the deployment of a modern nuclear computational chain which provides, besides advanced 3D tools for coupled neutronics/ thermal-hydraulics full core calculations, also appropriate tools for the generation of multi-group cross sections and Monte Carlo models for the verification of the individual calculational steps. This computational chain shall primarily be deployed for light water reactors (LWR), but should beyond that also be applicable for innovative reactor concepts. Thus, validation on computational benchmarks and critical experiments was of paramount importance. Finally, appropriate methods for uncertainty and sensitivity analysis were to be integrated into the computational framework, in order to assess and quantify the uncertainties due to insufficient knowledge of data, as well as due to methodological aspects.

Overview of on-line core monitoring system BEACON

International Nuclear Information System (INIS)

Dai Qing; Chen Xiaosong

2013-01-01

After more than 20 years of development, key technologies embedded with such system have reached a certain degree of maturity among some foreign countries. However, domestically, there is no comparable system yet. Through in-depth research and analysis on the most widely used core monitoring system in the world, BEACON, it's hope that this will provide guidance on our independent development of the first core monitoring system in China. Excore detectors, core outlet thermocouples and incore movable detectors are used to provide measure data on the status of the core for BEACON. Under the assumption of nodal homogeneity, an effective fast group model is used to solve the diffusion equation, followed by core-wise interpolation by Green's function. Finally, reconstruction of a calculated core is fitted with measured data using the surface spline function. The most significant technological advances are core monitoring during unstable core conditions, the use of nodal expansion method to improve accuracy and the adoption of single point calibration to increase the period of recalibration for the whole core. (authors)

Monte Carlo simulation of core physics parameters of the Nigeria Research Reactor-1 (NIRR-1)

Energy Technology Data Exchange (ETDEWEB)

Jonah, S.A. [Reactor Engineering Section, Centre for Energy Research and Training, Ahmadu Bello University, Zaria, P.M.B. 1014 (Nigeria)], E-mail: jonahsa2001@yahoo.com; Liaw, J.R.; Matos, J.E. [RERTR Program, Nuclear Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

2007-12-15

The Monte Carlo N-Particle (MCNP) code, version 4C (MCNP4C) and a set of neutron cross-section data were used to develop an accurate three-dimensional computational model of the Nigeria Research Reactor-1 (NIRR-1). The geometry of the reactor core was modeled as closely as possible including the details of all the fuel elements, reactivity regulators, the control rod, all irradiation channels, and Be reflectors. The following reactor core physics parameters were calculated for the present highly enriched uranium (HEU) core: clean cold core excess reactivity ({rho}{sub ex}), control rod (CR) and shim worth, shut down margin (SDM), neutron flux distributions in the irradiation channels, reactivity feedback coefficients and the kinetics parameters. The HEU input model was validated by experimental data from the final safety analyses report (SAR). The model predicted various key neutronics parameters fairly accurately and the calculated thermal neutron fluxes in the irradiation channels agree with the values obtained by foil activation method. Results indicate that the established Monte Carlo model is an accurate representation of the NIRR-1 HEU core and will be used to perform feasibility for conversion to low enriched uranium (LEU)

In core system mapping reactor power distribution

International Nuclear Information System (INIS)

Yoriyaz, H.; Moreira, J.M.L.

1989-01-01

Based on the signals of SPND'S (Self Powered Neutron Detectors) distributed inside of a core, the spatial power distribution is obtained using the MAP program, developed in this work. The methodology applied in MAP program uses a least mean square technique to calculate expansion coefficients that depend on the SPND'S signals. The final power or neutron flux distribution is obtained by a combination of certains functions or expansion modes that are provided from diffusion calculation with the CITATION code. The MAP program is written in PASCAL language and will be used in IEA-R1 reactor for assisting its operation. (author) [pt

MIMI: Multimodality, Multiresource, Information Integration Environment for Biomedical Core Facilities

OpenAIRE

Szymanski, Jacek; Wilson, David L.; Zhang, Guo-Qiang

2007-01-01

The rapid expansion of biomedical research has brought substantial scientific and administrative data management challenges to modern core facilities. Scientifically, a core facility must be able to manage experimental workflow and the corresponding set of large and complex scientific data. It must also disseminate experimental data to relevant researchers in a secure and expedient manner that facilitates collaboration and provides support for data interpretation and analysis. Administrativel...

MTR (Materials Testing Reactors) cores fuel management. Application of a low enrichment reactor for the equilibrium and transitory core calculation

International Nuclear Information System (INIS)

Relloso, J.M.

1990-01-01

This work describes a methodology to define the equilibrium core and a MTR (Materials Testing Reactors) type reactor's fuel management upon multiple boundary conditions, such as: end cycle and permitted maximum reactivities, burn-up extraction and maximun number of movements by rechange. The methodology proposed allows to determine the best options through conceptual relations, prior to a detailed calculation with the core code, reducing the test number with these codes and minimizing in this way CPU cost. The way to better systematized search of transient cores from the first one to the equilibrium one is presented. (Author) [es

Reactivity monitoring for safety purposes on the UK prototype fast reactor

International Nuclear Information System (INIS)

Lord, D.J.; Wilkes, D.J.

1987-01-01

The small size and high rating of the liquid metal cooled fast breeder reactor (LMFBR) make the provision of safety related instrumentation for individual subassemblies both difficult and expensive. Global monitoring of the core is thus very attractive. Reactivity monitoring is an important part of such global monitoring. Reactivity monitoring on a short timescale (a few seconds) is used on the UK Prototype Fast Reactor (PFR) as a trip parameter and long-term reactivity monitoring is being developed as a means of providing early warning of slowly developing faults. Results are presented from PFR to demonstrate the capabilities of reactivity monitoring in an operational fast reactor power station. (author)

Trends vs. reactor size of passive reactivity shutdown and control performance

International Nuclear Information System (INIS)

Wade, D.C.; Fujita, E.K.

1987-01-01

For LMR concepts, the goal of passive reactivity shutdown has been approached in the US by designing the reactors for favorable relationships among the power, power/flow, and inlet temperature coefficients of reactivity, for high internal conversion ratio (yielding small burnup control swing), and for a primary pump coastdown time appropriately matched to the delayed neutron hold back of power decay upon negative reactivity input. The use of sodium bonded metallic fuel pins has facilitated the achievement of the massive shutdown design goals as a consequence of their high thermal conductivity and high effective heavy metal density. Alternately, core designs based on derated oxide pins may be able to achieve the passive shutdown features at the cost of larger core volume and increased initial fissile inventory. For LMR concepts, the passive decay heat removal goal of inherent safety has been approached in US designs by use of pool layouts, larger surface to volume ratio of the reactor vessel with natural draft air cooling of the vessel surface, elevations and redans which promote natural circulation through the core, and thermal mass of the pool contents sufficient to absorb that initial transient decay heat which exceeds the natural draft air cooling capacity. This paper describes current US ''inherently safe'' reactor design

Measurement of xenon reactivity in the reactor of the nuclear ship 'MUTSU'

International Nuclear Information System (INIS)

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

1993-01-01

This report deals with the measurement of reactivity changes caused by the increase and decrease of xenon concentration in the reactor core of the nuclear ship 'MUTSU' after a change from long-term operation at 70 % to zero power. The change in xenon reactivity was compensated by control-rod movements and the compensated reactivity was measured using a digital reactivity meter. The xenon override peak was recognized five and half hours after the start of power reduction. The equilibrium and peak reactivities of xenon were estimated by reading the initial and peak values of a theoretical curve which was fitted to the measured variation in xenon reactivity. The xenon reactivity results obtained by the present method can be considered to be accurate since no control-rod worth data were used and the measured quantity was the reactivity itself. (author)

Probabilistic method for evaluating reactivity margin of nuclear reactors

International Nuclear Information System (INIS)

Kaneko, Yoshihiko

1984-01-01

A probabilistic method is proposed that will permit in the design stage to estimate quantitatively the likelihood with which any or all design criteria applicable to a nuclear reactor are actually satisfied after its construction. The method is trially applied to the core reactivity balance problem of the experimental Very High Temperature Reactor, and calculations are performed on the probability with which a design study core will, upon construction, satisfy design criteria concerning (a) one rod stuck and (b) startup margin. The method should prove useful in making engineering judgments before approving reactor core design. (author)

RA reactor reactivity changes before refurbishment - Task 3.08/02; Zadatak 3.08/02 - Promene reaktivnosti reaktora RA do remonta

Energy Technology Data Exchange (ETDEWEB)

Dobrosavljevic, N; Strugar, P; Stamenkovic, S [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

1963-12-15

From the the end of 1959, when the RA reactor started operation until January 1963 reactor was operated with the initial fuel batch of 56 fuel channels. After 310 MWd 68 fuel channels were added to the reactor core, and after further 357 MWd the core was filled up to the maximum of 88 fuel channels. Basic reactor parameters were systematically measured during two years of operation. This report covers the measurements concerned directly with the reactor operation: calibration of the control rods and their reactivity worths during operation, determining the total built-in reactivity excess and its change during burnup, determination of reactivity dependence on the temperature, xenon effect in the core.

Central Reactivity Measurements on Assemblies 1 and 3 of the Fast Reactor FR0

International Nuclear Information System (INIS)

Londen, S.O.

1966-01-01

The reactivity effects of small samples of various materials have been measured, by the period method at the core centre of Assemblies 1 and 3 of the fast zero power reactor FR0. For some materials the reactivity change as a function of sample size has also been determined experimentally. The core of Assembly 1 consisted only of uranium enriched to 20 % whereas the core of Assembly 3 was diluted with 30 % graphite. The results have been compared with calculated values obtained with a second-order transport-theoretical perturbation model and using differently shielded cross sections depending upon sample size. Qualitative agreement has generally been found, although discrepancies still exist. The spectrum perturbation caused by the experimental arrangement has been analyzed and found to be rather important

Central Reactivity Measurements on Assemblies 1 and 3 of the Fast Reactor FR0

Energy Technology Data Exchange (ETDEWEB)

Londen, S O

1966-01-15

The reactivity effects of small samples of various materials have been measured, by the period method at the core centre of Assemblies 1 and 3 of the fast zero power reactor FR0. For some materials the reactivity change as a function of sample size has also been determined experimentally. The core of Assembly 1 consisted only of uranium enriched to 20 % whereas the core of Assembly 3 was diluted with 30 % graphite. The results have been compared with calculated values obtained with a second-order transport-theoretical perturbation model and using differently shielded cross sections depending upon sample size. Qualitative agreement has generally been found, although discrepancies still exist. The spectrum perturbation caused by the experimental arrangement has been analyzed and found to be rather important.

Neutronics simulations on hypothetical power excursion and possible core melt scenarios in CANDU6

International Nuclear Information System (INIS)

Kim, Yonghee

2015-01-01

LOCA (Loss of coolant accident) is an outstanding safety issue in the CANDU reactor system since the coolant void reactivity is strongly positive. To deal with the LOCA, the CANDU systems are equipped with specially designed quickly-acting secondary shutdown system. Nevertheless, the so-called design-extended conditions are requested to be taken into account in the safety analysis for nuclear reactor systems after the Fukushima accident. As a DEC scenario, the worst accident situation in a CANDU reactor system is a unprotected LOCA, which is supposed to lead to a power excursion and possibly a core melt-down. In this work, the hypothetical unprotected LOCA scenario is simulated in view of the power excursion and fuel temperature changes by using a simplified point-kinetics (PK) model accounting for the fuel temperature change. In the PK model, the core reactivity is assumed to be affected by a large break LOCA and the fuel temperature is simulated to account for the Doppler effect. In addition, unlike the conventional PK simulation, we have also considered the Xe-I model to evaluate the impact of Xe during the LOCA. Also, we tried to simulate the fuel and core melt-down scenario in terms of the reactivity through a series of neutronics calculations for hypothetical core conditions. In case of a power excursion and possible fuel melt-down situation, the reactor system behavior is very uncertain. In this work, we tried to understand the impacts of fuel melt and relocation within the pressure vessel on the core reactivity and failure of pressure and calandria tubes. (author)

Improvement of SSR core design for ABWR-II

International Nuclear Information System (INIS)

Moriwaki, Masanao; Aoyama, Motoo; Okada, Hiroyuki; Kitamura, Hideya; Sakurada, Koichi; Tanabe, Akira

2003-01-01

In order to enhance the spectral shift effect in the ABWR-II reactor, a novel core design to bring out better performance of spectral shift rods (SSRs) is studied. The SSR is a new type of water rod, in which the water level develops naturally during operation and changes according to the coolant flow rate through the channel. By using the SSR, the average moderator density, which is directly related to core reactivity, can be controlled over a wide range by the core flow rate. In the new SSR core design, two types of SSR bundles, in which settings for the SSR water levels are different, are utilized and loaded according to flow distribution in the core. This two-region SSR core design allows wide variation in the average SSR water level, thus improving fuel economy. Enhancement of SSR function in the two-region SSR core increases the uranium saving factor by about 25%, from the 6% of the conventional uniform SSR core to about 8%. (author)

Analysis of LOF/TOP excursions in the SNR-300 Zielkern core with a stress controlled rip extension after failure

International Nuclear Information System (INIS)

Royl, P.

1987-07-01

An unprotected loss of flow (ULOF) accident can result in energetic power excursions due to reactivity effects from in-pile fuel motion and compaction after a near midplane failure of fuel pins in unvoided regions of the core (loss of flow driven transient overpower or LOF/TOP excursions). These excursions are strongly mitigated by a propagation of the failure rip, which reduces the fuel compaction and enhances fuel sweep-out. A simple model has been introduced into the SAS3D code in which this propagation is restricted to the expansion of the void zone that is firmed in the coolant channel due to thermal interactions of fuel and sodium (FCI) after failure. The rip propagation is demonstrated in this report in all investigated cases to be a strong mitigator that will limit the fuel compaction effects in the region where they originate

Investigating heavy water zero power reactors with a new core configuration based on experiment and calculation results

Energy Technology Data Exchange (ETDEWEB)

Nasrazadani, Zahra; Salimi, Raana; Askari, Afrooz; Khorsandi, Jamshid; Mirvakili, Mohammad; Mashayekh, Mohammad [Reactor Research School, Nuclear Science and Technology Research Institute, Atomic Energy Organization of Iran, Esfahan (Iran, Islamic Republic of)

2017-02-15

The heavy water zero power reactor (HWZPR), which is a critical assembly with a maximum power of 100 W, can be used in different lattice pitches. The last change of core configuration was from a lattice pitch of 18-20 cm. Based on regulations, prior to the first operation of the reactor, a new core was simulated with MCNP (Monte Carlo N-Particle)-4C and WIMS (Winfrith Improved Multigroup Scheme)-CITATON codes. To investigate the criticality of this core, the effective multiplication factor (Keff) versus heavy water level, and the critical water level were calculated. Then, for safety considerations, the reactivity worth of D2O, the reactivity worth of safety and control rods, and temperature reactivity coefficients for the fuel and the moderator, were calculated. The results show that the relevant criteria in the safety analysis report were satisfied in the new core. Therefore, with the permission of the reactor safety committee, the first criticality operation was conducted, and important physical parameters were measured experimentally. The results were compared with the corresponding values in the original core.

Device for protecting deformations of reactor cores

International Nuclear Information System (INIS)

Kato, Yasuyoshi; Urushihara, Hiroshi.

1975-01-01

Object: To provide a fluid pressure cylinder, which is operated according to change in temperature of coolant for a reactor to restrain or release a core, to simply and effectively protect deformation of the core. Structure: A closed fluid pressure cylinder interiorly filled with suitable fluid is disposed in peripherally equally spaced relation in an annular space between a core barrel of a reactor and a reactor vessel. A piston is mounted in fluid-tight fashion in a plurality of piston openings made in the cylinder, the piston being slidably moved according to expansion and contraction of the fluid filled in the cylinder. The piston has a movable frame mounted at the foremost end thereof, the movable frame being moved integral with the piston, and the surface opposite the mount thereof biasing the outermost peripheral surface of the core. (Kamimura, M.)

Optimization programs for reactor core fuel loading exhibiting reduced neutron leakage

International Nuclear Information System (INIS)

Darilek, P.

1991-01-01

The program MAXIM was developed for the optimization of the fuel loading of WWER-440 reactors. It enables the reactor core reactivity to be maximized by modifying the arrangement of the fuel assemblies. The procedure is divided into three steps. The first step includes the passage from the three-dimensional model of the reactor core to the two-dimensional model. In the second step, the solution to the problem is sought assuming that the multiplying properties, or the reactivity in the zones of the core, vary continuously. In the third step, parameters of actual fuel assemblies are inserted in the ''continuous'' solution obtained. Combined with the program PROPAL for a detailed refinement of the loading, the program MAXIM forms a basis for the development of programs for the optimization of fuel loading with burnable poisons. (Z.M.). 16 refs

Core clamping device for a nuclear reactor

International Nuclear Information System (INIS)

Guenther, R.W.

1974-01-01

The core clamping device for a fast neutron reactor includes clamps to support the fuel zone against the pressure vessel. The clamps are arranged around the circumference of the core. They consist of torsion bars arranged parallel at some distance around the core with lever arms attached to the ends whose force is directed in the opposite direction, pressing against the wall of the pressure vessel. The lever arms and pressure plates also actuated by the ends of the torsion bars transfer the stress, the pressure plates acting upon the fuel elements or fuel assemblies. Coupling between the ends of the torsion bars and the pressure plates is achieved by end carrier plates directly attached to the torsion bars and radially movable. This clamping device follows the thermal expansions of the core, allows specific elements to be disengaged in sections and saves space between the core and the neutron reflectors. (DG) [de

Characterization and reactivity of soot from fast pyrolysis of lignocellulosic compounds and monolignols

DEFF Research Database (Denmark)

Trubetskaya, Anna; Brown, Avery; Tompsett, Geoffrey

2018-01-01

spectroscopy. The CO2 reactivity of soot was investigated by thermogravimetric analysis. Soot from cellulose was more reactive than soot produced from extractives, lignin and monolignols. Soot reactivity was correlated with the separation distances between adjacent graphene layers, as measured using...... transmission electron microscopy. Particle size, free radical concentration, differences in a degree of curvature and multi-core structures influenced the soot reactivity less than the interlayer separation distances. Soot yield was correlated with the lignin content of the feedstock. The selection...... of the extraction solvent had a strong influence on the soot reactivity. The Soxhlet extraction of softwood and wheat straw lignin soot using methanol decreased the soot reactivity, whereas acetone extraction had only a modest effect....

A reactive empirical bond order (REBO) potential for hydrocarbon-oxygen interactions

International Nuclear Information System (INIS)

Ni, Boris; Lee, Ki-Ho; Sinnott, Susan B

2004-01-01

The expansion of the second-generation reactive empirical bond order (REBO) potential for hydrocarbons, as parametrized by Brenner and co-workers, to include oxygen is presented. This involves the explicit inclusion of C-O, H-O, and O-O interactions to the existing C-C, C-H, and H-H interactions in the REBO potential. The details of the expansion, including all parameters, are given. The new, expanded potential is then applied to the study of the structure and chemical stability of several molecules and polymer chains, and to modelling chemical reactions among a series of molecules, within classical molecular dynamics simulations

Analysis of the SPERT III E-core experiment using the EUREKA-2 code

International Nuclear Information System (INIS)

Harami, Taikan; Uemura, Mutsumi; Ohnishi, Nobuaki

1986-09-01

EUREKA-2, a coupled nuclear thermal hydrodynamic kinetic code, was adapted for the testing of models and methods. Code evaluations were made with the reactivity addition experiments of the SPERT III E-Core, a slightly enriched oxide core. The code was tested for non damaging power excursions including a wide range of initial operating conditions, such as cold-startup, hot-startup, hot-standby and operating-power initial conditions. Comparisons resulted in a good agreement within the experimental errors between calculated and experimental power, energy, reactivity and clad surface temperature. (author)

Experimental study on reactivity measurement in thermal reactor by polarity correlation method

International Nuclear Information System (INIS)

Yasuda, Hideshi

1977-11-01

Experimental study on the polarity correlation method for measuring the reactivity of a thermal reactor, especially the one possessing long prompt neutron lifetime such as graphite on heavy water moderated core, is reported. The techniques of reactor kinetics experiment are briefly reviewed, which are classified in two groups, one characterized by artificial disturbance to a reactor and the other by natural fluctuation inherent in a reactor. The fluctuation phenomena of neutron count rate are explained using F. de Hoffman's stochastic method, and correlation functions for the neutron count rate fluctuation are shown. The experimental results by polarity correlation method applied to the β/l measurements in both graphite-moderated SHE core and light water-moderated JMTRC and JRR-4 cores, and also to the measurement of SHE shut down reactivity margin are presented. The measured values were in good agreement with those by a pulsed neutron method in the reactivity range from critical to -12 dollars. The conditional polarity correlation experiments in SHE at -20 cent and -100 cent are demonstrated. The prompt neutron decay constants agreed with those obtained by the polarity correlation experiments. The results of experiments measuring large negative reactivity of -52 dollars of SHE by pulsed neutron, rod drop and source multiplication methods are given. Also it is concluded that the polarity and conditional polarity correlation methods are sufficiently applicable to noise analysis of a low power thermal reactor with long prompt neutron lifetime. (Nakai, Y.)

Stand-alone core sensitivity and uncertainty analysis of ALFRED from Monte Carlo simulations

International Nuclear Information System (INIS)

Pérez-Valseca, A.-D.; Espinosa-Paredes, G.; François, J.L.; Vázquez Rodríguez, A.; Martín-del-Campo, C.

2017-01-01

Highlights: • Methodology based on Monte Carlo simulation. • Sensitivity analysis of Lead Fast Reactor (LFR). • Uncertainty and regression analysis of LFR. • 10% change in the core inlet flow, the response in thermal power change is 0.58%. • 2.5% change in the inlet lead temperature the response is 1.87% in power. - Abstract: The aim of this paper is the sensitivity and uncertainty analysis of a Lead-Cooled Fast Reactor (LFR) based on Monte Carlo simulation of sizes up to 2000. The methodology developed in this work considers the uncertainty of sensitivities and uncertainty of output variables due to a single-input-variable variation. The Advanced Lead fast Reactor European Demonstrator (ALFRED) is analyzed to determine the behavior of the essential parameters due to effects of mass flow and temperature of liquid lead. The ALFRED core mathematical model developed in this work is fully transient, which takes into account the heat transfer in an annular fuel pellet design, the thermo-fluid in the core, and the neutronic processes, which are modeled with point kinetic with feedback fuel temperature and expansion effects. The sensitivity evaluated in terms of the relative standard deviation (RSD) showed that for 10% change in the core inlet flow, the response in thermal power change is 0.58%, and for 2.5% change in the inlet lead temperature is 1.87%. The regression analysis with mass flow rate as the predictor variable showed statistically valid cubic correlations for neutron flux and linear relationship neutron flux as a function of the lead temperature. No statistically valid correlation was observed for the reactivity as a function of the mass flow rate and for the lead temperature. These correlations are useful for the study, analysis, and design of any LFR.

Some characteristics of two heterogeneous cores and their experimental confirmation

International Nuclear Information System (INIS)

Giese, H.; Henneges, G.; Pilate, S.

1979-01-01

Heterogeneous core geometries have been investigated at the zero power facility ZEBRA with the objective of developing an improved understanding of the basic physics parameters of future non-conventional fast breeder reactors. Three assemblies were investigated, comprising similar fissile volumes and numbers of in-core fertile elements. First, BZC, a Salt-and-Pepper Core with fertile elements arranged in groups of sizes ranging from about one to eight power reactor subassemblies. Second, BZD, a Single Annular Core with the fertile elements collected to form a large central island of about 95 cm diameter, equivalent to about 36 subassemblies. Third, BZD/1A, a modified version of assembly BZD with a central island diameter of about 67 cm and an additional thin breeder ring. All assemblies used 24% Pu/Pu + U enriched fuel, with a core height of about 90 cm and a core diameter of about 200 cm. Extensive sodium voiding, flux-tilt and reactivity-interaction measurements have been carried out. The results indicate that relative to homogeneous designs a Single Annular arrangement yields a more marked improvement in sodium voiding performance than a Salt-and-Pepper arrangement, at the expense, however, of an increased sensitivity of the flux distribution to local reactivity perturbations. First analysis-level calculations on sodium-void yield C/E values with a significantly wider dispersion than in previous homogeneous assemblies. (author)

Medicaid Expansion And Grant Funding Increases Helped Improve Community Health Center Capacity.

Science.gov (United States)

Han, Xinxin; Luo, Qian; Ku, Leighton

2017-01-01

Through the expansion of Medicaid eligibility and increases in core federal grant funding, the Affordable Care Act (ACA) sought to increase the capacity of community health centers to provide primary care to low-income populations. We examined the effects of the ACA Medicaid expansion and changes in federal grant levels on the centers' numbers of patients, percentages of patients by type of insurance, and numbers of visits from 2012 to 2015. In the period after expansion (2014-15), health centers in expansion states had a 5 percent higher total patient volume, larger shares of Medicaid patients, smaller shares of uninsured patients, and increases in overall visits and mental health visits, compared to centers in nonexpansion states. Increases in federal grant funding levels were associated with increases in numbers of patients and of overall, medical, and preventive service visits. If federal grant levels are not sustained after 2017, there could be marked reductions in health center capacity in both expansion and nonexpansion states. Project HOPE—The People-to-People Health Foundation, Inc.

A digital instrument for reactivity measurements in a nuclear reactor

International Nuclear Information System (INIS)

Chwaszczewski, S.

1979-01-01

An instrument for digital determination of the reactivity in nuclear reactors is described. It is based on the CAMAC standard apparatus, suitable for the use of pulse or current type neutron detectors and operates with prompt response and an output signal proportional to the core neutron flux. The measured data of neutron flux and reactivity can be registered by a digital display unit, an indicator, or, by request of the operator, a paper type punch. The algorithms used for reactivity calculation are considered and the results of numerical studies on those algorithms are discussed. The instrument has been used for determining the reactivity of the control elements in the fast-thermal assembly ANNA and in the research reactor MARIA. Some results of these measurements are given. (author)

Experimental estimation of moderator temperature coefficient of reactivity of the IPEN/MB-01 research reactor

Energy Technology Data Exchange (ETDEWEB)

Silva, Rubens C. da; Bitelli, Ulysses D.; Mura, Luiz Ernesto C., E-mail: rubensrcs@usp.br, E-mail: ubitelli@ipen.br, E-mail: credidiomura@gmail.com [Universidade de Sao Paulo (PNV/POLI/USP), SP (Brazil). Arquitetura Naval e Departamento de Engenharia Oceanica; Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2017-07-01

The aim of this article is to present the procedure for the experimental estimation of the Moderator Temperature Coefficient of Reactivity of the IPEN/MB-01 Research Reactor, a parameter that has an important role in the physics and the control operations of any reactor facility. At the experiment, the IPEN/MB-01 reactor went critical at the power of 1W (1% of its total power), and whose core configuration was 28 x 26 rectangular array of UO{sub 2} fuel rods, inside a light water (moderator) tank. In addition, there was a heavy water (D{sub 2}O) reflector installed in the West side of the core to obtain an adequate neutron reflection along the experiment. The moderator temperature was increased in steps of 4 °C, and the measurement of the mean moderator temperature was acquired using twelve calibrated thermocouples, placed around the reactor core. As a result, the mean value of -4.81 pcm/°C was obtained for such coefficient. The curves of ρ(T) (Reactivity x Temperature) and α{sup M}{sub T}(T)(Moderator Temperature Coefficient of Reactivity x Temperature) were developed using data from an experimental measurement of the integral reactivity curves through the Stable Period and Inverse Kinetics Methods, that was carried out at the reactor with the same core configuration. Such curves were compared and showed a very similar behavior between them. (author)

On the area expansion of magnetic flux tubes in solar active regions

Energy Technology Data Exchange (ETDEWEB)

Dudík, Jaroslav [DAMTP, CMS, University of Cambridge, Wilberforce Road, Cambridge CB3 0WA (United Kingdom); Dzifčáková, Elena [Astronomical Institute of the Academy of Sciences of the Czech Republic, Fričova 298, 251 65 Ondřejov (Czech Republic); Cirtain, Jonathan W., E-mail: J.Dudik@damtp.cam.ac.uk, E-mail: elena@asu.cas.cz [NASA Marshall Space Flight Center, VP 62, Huntsville, AL 35812 (United States)

2014-11-20

We calculated the three-dimensional (3D) distribution of the area expansion factors in a potential magnetic field, extrapolated from the high-resolution Hinode/SOT magnetogram of the quiescent active region NOAA 11482. Retaining only closed loops within the computational box, we show that the distribution of area expansion factors show significant structure. Loop-like structures characterized by locally lower values of the expansion factor are embedded in a smooth background. These loop-like flux tubes have squashed cross-sections and expand with height. The distribution of the expansion factors show an overall increase with height, allowing an active region core characterized by low values of the expansion factor to be distinguished. The area expansion factors obtained from extrapolation of the Solar Optical Telescope magnetogram are compared to those obtained from an approximation of the observed magnetogram by a series of 134 submerged charges. This approximation retains the general flux distribution in the observed magnetogram, but removes the small-scale structure in both the approximated magnetogram and the 3D distribution of the area expansion factors. We argue that the structuring of the expansion factor can be a significant ingredient in producing the observed structuring of the solar corona. However, due to the potential approximation used, these results may not be applicable to loops exhibiting twist or to active regions producing significant flares.

Modeling the reactor core of MNSR to simulate its dynamic behavior using the code PARET

International Nuclear Information System (INIS)

Hainoun, A.; Alhabet, F.

2004-02-01

Using the computer code PARET the core of the MNSR reactor was modelled and the neutronics and thermal hydraulic behaviour of the reactor core for the steady state and selected transients, that deal with step change of reactivity including control rod withdraw starting from steady state at various low power level, were simulated. For this purpose a PARET input model for the core of MNSR reactor has been developed enabling the simulation of neutron kinetic and thermal hydraulic of reactor core including reactivity feedback effects. The neutron kinetic model depends on the point kinetic with 15 groups delayed neutrons including photo neutrons of beryllium reflector. In this regard the effect of photo neutron on the dynamic behaviour has been analysed through two additional calculation. In the first the yield of photo neutrons was neglected completely and in the second its share was added to the sixth group of delayed neutrons. In the thermal hydraulic model the fuel elements with their cooling channels were distributed to 4 different groups with various radial power factors. The pressure lose factors for friction, flow direction change, expansion and contraction were estimated using suitable approaches. The post calculations of the relative neutron flux change and core average temperature were found to be consistent with the experimental measurements. Furthermore, the simulation has indicated the influence of photo neutrons of the Beryllium reflector on the neutron flux behaviour. For the reliability of the results sensitivity analysis was carried out to consider the uncertainty in some important parameters like temperature feedback coefficient and flow velocity. On the other hand the application of PARET in simulation of void formation in the subcooled boiling regime were tested. The calculation indicates the capability of PARET in modelling this phenomenon. However, big discrepancy between calculation results and measurement of axial void distribution were observed

Digital reactivity meter construction based on PC

International Nuclear Information System (INIS)

Yusi-Eko-Yulianto; Kristedjo-Kurnianto

2003-01-01

The reactivitymeter is a core reactivity measuring equipment, which inform the reactor operator the neutron flux development in the core. This digital reactivitymeter is needed to replace analog reactivitymeter, whenever it fails in the future. The replacement of thus reactivitymeter can keep the continuation of reactor operation. The digital reactivitymeter is constructed by using the digital signal processing and computer. Thus real time signal processing is displayed on the monitor graphically. This reactivitymeter has been tested in RSG-GAS and perform a good work. This performance is worthy to use this digital reactivitymeter for RSG-GAS operation

Optimal burnable poison utilization in PWR core reload design

International Nuclear Information System (INIS)

Downar, T.J.

1986-01-01

A method was developed for determining the optimal distribution and depletion of burnable poisons in a Pressurized Water Reactor core. The well-known Haling depletion technique is used to achieve the end-of-cycle core state where the fuel assembly arrangement is configured in the absence of all control poison. The soluble and burnable poison required to control the core reactivity and power distribution are solved for as unknown variables while step depleting the cycle in reverse with a target power distribution. The method was implemented in the NRC approved licensing code SIMULATE

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

International Nuclear Information System (INIS)

Shin, Hyeong-Ki

1999-01-01

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

An approach of SFR safety study through the most penalizing sodium void reactivity - 105

International Nuclear Information System (INIS)

Tiberi, V.; Ivanov, E.; Pignet, S.

2010-01-01

Sodium void reactivity effects can affect the plant safety significantly during accidental transients. Accordingly, they have to be accurately investigated for any new sodium cooled fast reactor concept, even if a fuel with a melting point lower than the sodium boiling temperature is adopted. Thus all new reactor concepts should be compared to each - others adopting the value of the maximal possible sodium void reactivity as a discrimination parameter. However, taking into account that the sodium void worth is spatially depended, it is not evident which volume could be voided in order to obtain the maximal reactivity increase. Typically the complete active core voiding (zones initially loaded with 235 U or 239 Pu) is taken into account. This paper summarizes the extensive work carried-out in the IRSN to investigate the sodium-void reactivity spatial profiles of a fast sodium-cooled reactor core in the aim of defining a methodology to search for the area where the void contribution to the reactivity is strictly positive. Perturbation theory design approach available in the ERANOS 2.1 code has been adopted to evaluate the 'area of positive void worth'. To do that, the code has been previously validated against experimental based benchmarks (IRPhEP) and reference calculations. The evaluation of the absolute values of reactivity profiles can be improved later-on adopting more sophisticated methodologies to perform more accurate calculations of the sample with the voided area determined adopting the rough procedure described here. It has been demonstrated that even the non-reference way of ERANOS calculations could be used to provide the basis for different core concepts inter-comparison. (authors)

Performance testing of a mixed TRIGA core

Energy Technology Data Exchange (ETDEWEB)

Schumacher, R F; Godsey, T A; Feltz, D E; Randall, J D [Texas A and M University (United States)

1974-07-01

The major operational problem experienced by the Nuclear Science Center Reactor at Texas A and M University is full burnup. With two shift operation caused by the high utilization of the facility, the reactor is operated more than 100 megawatt days per year. The solution chosen for this problem was conversion to FLIP fuel. Since funds were not available to load an entire FLIP core, a mixed core comprised of approximately one third FLIP fuel located in the central region was designed. The design core was loaded and went critical on July 1, 1973. The results of the following measurements on the mixed core are presented: Determination of Rod worths; Measurement of Reactivity Effects; Determination of Flux values; Measurement of Fuel temperatures; Preliminary Fuel Burnup Rate; Pulsing Calibration. (author)

RAtional Mapping (RAM) of in-core data

International Nuclear Information System (INIS)

Bonalumi, R.A.; Kherani, N.P.

1983-01-01

The paper describes and demonstrates a unique processing of in-core flux detector data, such that the detailed in-core power distribution can be derived with great accuracy by combining a specially 'smoothed-out' set of in-core data with neutron diffusion theory. RAM is designed in such a way that erratic detector signals are recognized very efficiently and can be eliminated from the experimental data set: this is achieved by modal expansion of the difference between theoretical fluxes and experimental fluxes at the detector sites. Sensitivity studies have shown that RAM is quite stable, does not absorb the 'wild' detector errors in the mapping procedure and results in mapped fluxes with errors about three times smaller than would be obtained by direct interpolation of detector readings

Proposal of a benchmark for core burnup calculations for a VVER-1000 reactor core

International Nuclear Information System (INIS)

Loetsch, T.; Khalimonchuk, V.; Kuchin, A.

2009-01-01

In the framework of a project supported by the German BMU the code DYN3D should be further validated and verified. During the work a lack of a benchmark on core burnup calculations for VVER-1000 reactors was noticed. Such a benchmark is useful for validating and verifying the whole package of codes and data libraries for reactor physics calculations including fuel assembly modelling, fuel assembly data preparation, few group data parametrisation and reactor core modelling. The benchmark proposed specifies the core loading patterns of burnup cycles for a VVER-1000 reactor core as well as a set of operational data such as load follow, boron concentration in the coolant, cycle length, measured reactivity coefficients and power density distributions. The reactor core characteristics chosen for comparison and the first results obtained during the work with the reactor physics code DYN3D are presented. This work presents the continuation of efforts of the projects mentioned to estimate the accuracy of calculated characteristics of VVER-1000 reactor cores. In addition, the codes used for reactor physics calculations of safety related reactor core characteristics should be validated and verified for the cases in which they are to be used. This is significant for safety related evaluations and assessments carried out in the framework of licensing and supervision procedures in the field of reactor physics. (authors)

Analysis of multiple failure accident scenarios for development of probabilistic safety assessment model for KALIMER-600

International Nuclear Information System (INIS)

Kim, T.W.; Suk, S.D.; Chang, W.P.; Kwon, Y.M.; Jeong, H.Y.; Lee, Y.B.; Ha, K.S.; Kim, S.J.

2009-01-01

A sodium-cooled fast reactor (SFR), KALIMER-600, is under development at KAERI. Its fuel is the metal fuel of U-TRU-Zr and it uses sodium as coolant. Its advantages are found in the aspects of an excellent uranium resource utilization, inherent safety features, and nonproliferation. The probabilistic safety assessment (PSA) will be one of the initiating subjects for designing it from the aspects of a risk informed design (RID) as well as a technology-neutral licensing (TNL). The core damage is defined as coolant voiding, fuel melting, or cladding damage. Accident scenarios which lead to the core damage should be identified for the development of a Level-1 PSA model. The SSC-K computer code is used to identify the conditions which lead to core damage. KALIMER-600 has passive safety features such as passive shutdown functions, passive pump coast-down features, and passive decay heat removal systems. It has inherent reactivity feedback effects such as Doppler, sodium void, core axial expansion, control rod axial expansion, core radial expansion, etc. The accidents which are analyzed are the multiple failure accidents such as an unprotected transient overpower, a loss of flow, and a loss of heat sink events with degraded safety systems or functions. The safety functions to be considered here are a reactor trip, inherent reactivity feedback features, the pump coast-down, and the passive decay heat removal. (author)

Sensitivity analysis of power excursion in RSG-GAS reactor due to reactivity insertion

International Nuclear Information System (INIS)

Pinem, Surian; Sembiring, Tagor Malem

2002-01-01

Reactor kinetics has a very important role in reactor operation safety and nuclear reactor control. One of the important aspects in reactor kinetics is power behavior as function of time due to chain reaction in the core. The calculation was performed using kinetic equation and feedback reactivity and evaluated using static power coefficient. Analysis was performed for oxide 250 g, silicide 250 g and silicide 300 g fuel elements with insertion of positive reactivity, negative reactivity and reactivity close to delay neutron fraction. The calculation of power excursion sensitivity showed that the insertion of 0,5 % Δk/k, in the fuel element of silicide 300 g is bigger 5 % than the one of oxide 250 g or silicide 250 g. If inserted by - 1,2 % Δk/k, there is no change among three fuel elements. Therefore, in kinetic point of view, it is showed there is no significant influence in the RSG-GAS reactor operation safety is the current core of oxide 250 g is converted to silicide 250 g or to silicide 300 g

IAEA Technical Meeting on Innovative Fast Reactor Designs with Enhanced Negative Reactivity Feedback Features. Presentations

International Nuclear Information System (INIS)

2012-01-01

The objective of the TM is to review and discuss the safety characteristics and the performances of the core of innovative fast reactor concepts, as well as to present the ongoing R&D activities in the area of core design and advanced simulation tools and methods for fast reactor core physics analysis. The focus is on fast spectrum cores optimized for actinide utilization and transmutation and, in particular, on core designs with enhanced negative reactivity feedback effects

Assessment of PWR safety with regard to disturbances due to reactivity changes

International Nuclear Information System (INIS)

Pernica, R.

1980-01-01

The steady state method is briefly described for reactivity disturbances assessment using steady state calculations for two sets of reactivity coefficients and four values of the thermal conductivity of the gap. The variations were processed of the limit values of reactivity being applied with the thermal conductivity of the gap between the fuel and the can. All calculations were performed for a reactor with four core zones exposed to different radial thermal stresses with different fuel element proportional stresses. The results are shown in graphs. (J.B.)

Investigation of the Buckling-Reactivity Conversion Coefficient using SRAC and MVP codes for UO2 Lattices in TCA experiments

International Nuclear Information System (INIS)

Le Dai Dien

2008-01-01

Benchmark experiments for International Reactor Physics Benchmark Experiments (IRPhE) Project carried out at TCA, the temperature effects on reactivity were studied for light water moderated and reflected UO 2 cores with/without soluble poisons. The buckling coefficient method using the measured critical water levels was proposed by Suzaki et al. The temperature dependence of buckling coefficient of reactivity and its variance by the core configurations of the benchmark experiments was investigated using SRAC and MVP calculations. From the calculations by SRAC as well as by MVP it is seen that the K-value can be taken as an average value only for each core with temperature changes which are considered as perturbation parameter. The difference between our calculations and benchmark results which uses constant K-value for all cores proves that the results depend on K-value and it play important role in defining reactivity effect using the water level worth method. (author)

Nuclear design and analysis report for KALIMER breakeven core conceptual design

International Nuclear Information System (INIS)

Kim, Sang Ji; Song, Hoon; Lee, Ki Bog; Chang, Jin Wook; Hong, Ser Gi; Kim, Young Gyun; Kim, Yeong Il

2002-04-01

During the phase 2 of LMR design technology development project, the breakeven core configuration was developed with the aim of the KALIMER self-sustaining with regard to the fissile material. The excess fissile material production is limited only to the extent of its own requirement for sustaining its planned power operation. The average breeding ratio is estimated to be 1.05 for the equilibrium core and the fissile plutonium gain per cycle is 13.9 kg. The nuclear performance characteristics as well as the reactivity coefficients have been analyzed so that the design evaluation in other activity areas can be made. In order to find out a realistic heavy metal flow evolution and investigate cycle-dependent nuclear performance parameter behaviors, the startup and transition cycle loading strategies are developed, followed by the startup core physics analysis. Driver fuel and blankets are assumed to be shuffled at the time of each reload. The startup core physics analysis has shown that the burnup reactivity swing, effective delayed neutron fraction, conversion ratio and peak linear heat generation rate at the startup core lead to an extreme of bounding physics data for safety analysis. As an outcome of this study, a whole spectrum of reactor life is first analyzed in detail for the KALIMER core. It is experienced that the startup core analysis deserves more attention than the current design practice, before the core configuration is finalized based on the equilibrium cycle analysis alone.

An fMRI study of nicotine-deprived smokers' reactivity to smoking cues during novel/exciting activity.

Directory of Open Access Journals (Sweden)

Xiaomeng Xu

Full Text Available Engaging in novel/exciting ("self-expanding" activities activates the mesolimbic dopamine pathway, a brain reward pathway also associated with the rewarding effects of nicotine. This suggests that self-expanding activities can potentially substitute for the reward from nicotine. We tested this model among nicotine-deprived smokers who, during fMRI scanning, played a series of two-player cooperative games with a relationship partner. Games were randomized in a 2 (self-expanding vs. not x 2 (cigarette cue present vs. absent design. Self-expansion conditions yielded significantly greater activation in a reward region (caudate than did non-self-expansion conditions. Moreover, when exposed to smoking cues during the self-expanding versus the non-self-expanding cooperative games, smokers showed less activation in a cigarette cue-reactivity region, a priori defined [temporo-parietal junction (TPJ] from a recent meta-analysis of cue-reactivity. In smoking cue conditions, increases in excitement associated with the self-expanding condition (versus the non-self-expanding condition were also negatively correlated with TPJ activation. These results support the idea that a self-expanding activity promoting reward activation attenuates cigarette cue-reactivity among nicotine-deprived smokers. Future research could focus on the parameters of self-expanding activities that produce this effect, as well as test the utility of self-expansion in clinical interventions for smoking cessation.

Use of thermodynamics and expansion in parts and machines; Utilisation de la thermodynamique de la detente dans les organes et les machines

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-12-31

This conference day was organized by the `thermodynamics` section of the French association of thermal engineers. This book of proceedings contains 11 papers entitled: `introduction talk`; `use of thermodynamics during expansion in valves in order to improve their behaviour`; `influence of the presence of impurities on the vapour condensation process during expansion inside a nozzle`; `excitations generated by flows during expansions`; `relief valves for refrigerating machineries`; `the influence of pressure losses in distribution systems over thermodynamic performances of the cryogenic pressure reducer and pneumatic engines`; `analysis of energy losses in turbine blades`; `humid vapor expansion in turbines`; `expansion of high temperature combustion gases in a turbine: influence of chemical reactivity`; `high expansion rate turbines`; `contribution of new calculation methods`. (J.S.)

Use of thermodynamics and expansion in parts and machines; Utilisation de la thermodynamique de la detente dans les organes et les machines

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-12-31

This conference day was organized by the `thermodynamics` section of the French association of thermal engineers. This book of proceedings contains 11 papers entitled: `introduction talk`; `use of thermodynamics during expansion in valves in order to improve their behaviour`; `influence of the presence of impurities on the vapour condensation process during expansion inside a nozzle`; `excitations generated by flows during expansions`; `relief valves for refrigerating machineries`; `the influence of pressure losses in distribution systems over thermodynamic performances of the cryogenic pressure reducer and pneumatic engines`; `analysis of energy losses in turbine blades`; `humid vapor expansion in turbines`; `expansion of high temperature combustion gases in a turbine: influence of chemical reactivity`; `high expansion rate turbines`; `contribution of new calculation methods`. (J.S.)

Benchmark on traveling wave fast reactor with negative reactivity feedback obtained with MCNPX code

International Nuclear Information System (INIS)

Gann, V.V.; Gann, A.V.

2012-01-01

This paper presents results of computer simulations of traveling wave fast reactor with negative reactivity feedback. The results were obtained using MCNPX code combined with CINDER90 subroutine for depletion calculations. We considered 1-D model of TWR containing 4 m long core made of mixture of 66 at. % 238 U and 34 at. % 10 B. Ignitor made of 235 U was located in the center of the core. Boron was included as imitator of structural in-core materials and coolant. Negative reactivity feedback was adjusted to reactor power of 500 MW. In this case two burning waves originated from the igniter and travel to the ends of the core during the following 40 years; coefficient of utilization of 238 U reached 80 %. Distribution of specific power in traveling wave, isotope concentration of fission products and actinides, neutron flux, fast neutron spectrum, specific activity were calculated. Data of the computer simulation is in qualitative agreement with theoretical results obtained in slow burning wave approximation

Neutronic analysis of LMFBRs during severe core disruptive accidents

International Nuclear Information System (INIS)

Tomlinson, E.T.

1979-01-01

A number of numerical experiments were performed to assess the validity of diffusion theory and various perturbation methods for calculating the reactivity state of a severely disrupted liquid metal cooled fast breeder reactor (LMFBR). The disrupted configurations correspond, in general, to phases through which an LMFBR core could pass during a core disruptive accident (CDA). Two-reactor models were chosen for this study, the two zone, homogeneous Clinch River Breeder Reactor and the Large Heterogeneous Reactor Design Study Core. The various phases were chosen to approximate the CDA results predicted by the safety analysis code SAS3D. The calculational methods investigated in this study include the eigenvalue difference technique based on both discrete ordinate transport theory and diffusion theory, first-order perturbation theory, exact perturbation theory, and a new hybrid perturbation theory. Selected cases were analyzed using Monte Carlo methods. It was found that in all cases, diffusion theory and perturbation theory yielded results for the change in reactivity that significantly disagreed with both the discrete ordinate and Monte Carlo results. These differences were, in most cases, in a nonconservative direction

Cell fiber-based three-dimensional culture system for highly efficient expansion of human induced pluripotent stem cells.

Science.gov (United States)

Ikeda, Kazuhiro; Nagata, Shogo; Okitsu, Teru; Takeuchi, Shoji

2017-06-06

Human pluripotent stem cells are a potentially powerful cellular resource for application in regenerative medicine. Because such applications require large numbers of human pluripotent stem cell-derived cells, a scalable culture system of human pluripotent stem cell needs to be developed. Several suspension culture systems for human pluripotent stem cell expansion exist; however, it is difficult to control the thickness of cell aggregations in these systems, leading to increased cell death likely caused by limited diffusion of gases and nutrients into the aggregations. Here, we describe a scalable culture system using the cell fiber technology for the expansion of human induced pluripotent stem (iPS) cells. The cells were encapsulated and cultured within the core region of core-shell hydrogel microfibers, resulting in the formation of rod-shaped or fiber-shaped cell aggregations with sustained thickness and high viability. By encapsulating the cells with type I collagen, we demonstrated a long-term culture of the cells by serial passaging at a high expansion rate (14-fold in four days) while retaining its pluripotency. Therefore, our culture system could be used for large-scale expansion of human pluripotent stem cells for use in regenerative medicine.

The whiteStar development project: Westinghouse's next generation core design simulator and core monitoring software to power the nuclear renaissance

International Nuclear Information System (INIS)

Boyd, W. A.; Mayhue, L. T.; Penkrot, V. S.; Zhang, B.

2009-01-01

The WhiteStar project has undertaken the development of the next generation core analysis and monitoring system for Westinghouse Electric Company. This on-going project focuses on the development of the ANC core simulator, BEACON core monitoring system and NEXUS nuclear data generation system. This system contains many functional upgrades to the ANC core simulator and BEACON core monitoring products as well as the release of the NEXUS family of codes. The NEXUS family of codes is an automated once-through cross section generation system designed for use in both PWR and BWR applications. ANC is a multi-dimensional nodal code for all nuclear core design calculations at a given condition. ANC predicts core reactivity, assembly power, rod power, detector thimble flux, and other relevant core characteristics. BEACON is an advanced core monitoring and support system which uses existing instrumentation data in conjunction with an analytical methodology for on-line generation and evaluation of 3D core power distributions. This new system is needed to design and monitor the Westinghouse AP1000 PWR. This paper describes provides an overview of the software system, software development methodologies used as well some initial results. (authors)

Monte Carlo analysis of Musashi TRIGA mark II reactor core

International Nuclear Information System (INIS)

Matsumoto, Tetsuo

1999-01-01

The analysis of the TRIGA-II core at the Musashi Institute of Technology Research Reactor (Musashi reactor, 100 kW) was performed by the three-dimensional continuous-energy Monte Carlo code (MCNP4A). Effective multiplication factors (k eff ) for the several fuel-loading patterns including the initial core criticality experiment, the fuel element and control rod reactivity worth as well as the neutron flux measurements were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated k eff overestimated the experimental data by about 1.0%Δk/k for both the initial core and the several fuel-loading arrangements. The calculated reactivity worths of control rod and fuel element agree well the measured ones within the uncertainties. The comparison of neutron flux distribution was consistent with the experimental ones which were measured by activation methods at the sample irradiation tubes. All in all, the agreement between the MCNP predictions and the experimentally determined values is good, which indicated that the Monte Carlo model is enough to simulate the Musashi TRIGA-II reactor core. (author)

Study of reactivity of fluidized bed nuclear reactor

International Nuclear Information System (INIS)

Rammsy, J.E.M.

1985-01-01

The reactor physics calculations of a 19 module Fluidized Bed Nuclear Reactor using Leopard and Odog codes are performed. The behaviour of the reactor was studied by calculating the reactivity of the reactor as a function of the parameters governing the operational and accidental conditions of the reactor. The effects of temperature, pressure, and vapor generation in the core on the reactivity are calculated. Also the start up behaviour of the reactor is analyzed. For the purpose of the study of a prototype research reactor, the calculations on a one module reactor have been performed. (Author) [pt

Dynamic behavior of homogeneous and heterogeneous LMFBR core-design concepts

International Nuclear Information System (INIS)

Chang, Y.I.; Henryson, H. II; Orechwa, Y.; Su, S.F.; Greenman, G.; Blomquist, R.

1981-01-01

The emphasis is placed on obtaining an understanding of the inherent difference between homogeneous and heterogeneous core configurations regarding neutronic characteristics related to the dynamic behavior. The space-time neutronic and thermal-hydraulic behavior was analyzed in detail for various core configurations by using the FX2-TH, a two-dimensional kinetics code with thermal-hydraulic feedback. In addition, the relationship between the flux tilt and the fundamental-to-first harmonic eigenvalue separation, and the sodium void reactivity in heterogeneous cores were also sutdied

A core management system for JRR-3

International Nuclear Information System (INIS)

Soyama, Kazuhiko; Tsuruta, Harumichi; Ichikawa, Hiroki; Nemoto, Hiroyuki.

1991-05-01

Japan Research Reactor No.3 (JRR-3) was upgraded to the thermal output with 20 MW by replacing the core, cooling system and utilization facilities. It is a water moderated and cooled, pool type reactor using 20% enriched U · Alx fuel. A core management system for JRR-3 has been made. This code system can manage of reactivity, power distribution and burn up in consideration of the position of control rod, fuel arrangement and operation pattern. This report is the user's manual of this code system. (author)

Effect of Strain on the Reactivity of Metal Surfaces

DEFF Research Database (Denmark)

Mavrikakis, Manos; Hammer, Bjørk; Nørskov, Jens Kehlet

1998-01-01

Self-consistent density functional calculations for the adsorption of O and CO, and the dissociation of CO on strained and unstrained Ru(0001) surfaces are used to show how strained metal surfaces have chemical properties that are significantly different from those of unstrained surfaces. Surface...... reactivity increases with lattice expansion, following a concurrent up-shift of the metal d states. Consequences for the catalytic activity of thin metal overlayers are discussed....

Value of large scale expansion of tumor infiltrating lymphocytes in a compartmentalised gas-permeable bag: interests for adoptive immunotherapy

Science.gov (United States)

2011-01-01

Background Adoptive cell therapy (ACT) has emerged as an effective treatment for patients with metastatic melanoma. However, there are several logistical and safety concerns associated with large-scale ex vivo expansion of tumour-specific T lymphocytes for widespread availability of ACT for cancer patients. To address these problems we developed a specific compartmentalised bag allowing efficient expansion of tumour-specific T lymphocytes in an easy handling, closed system. Methods Starting from lymph nodes from eight melanoma patients, we performed a side-by-side comparison of Tumour-Infiltrating Lymphocytes (TIL) produced after expansion in the compartmentalised bag versus TIL produced using the standard process in plates. Proliferation yield, viability, phenotype and IFNγ secretion were comparatively studied. Results We found no differences in proliferation yield and cell viability between both TIL production systems. Moreover, each of the cell products complied with our defined release criteria before being administered to the patient. The phenotype analysis indicated that the compartmentalised bag favours the expansion of CD8+ cells. Finally, we found that TIL stimulated in bags were enriched in reactive CD8+ T cells when co-cultured with the autologous melanoma cell line. Conclusions The stimulation of TIL with feeder cells in the specifically designed compartmentalised bag can advantageously replace the conventional protocol using plates. In particular, the higher expansion rate of reactive CD8+ T cells could have a significant impact for ACT. PMID:21575188

Evaluation of concrete pavements with materials-related distress : appendix B.

Science.gov (United States)

2010-02-02

An evaluation of cores sampled from six concrete pavements was performed. Factors contributing to pavement distress observed in the field were determined, including expansive alkali-silica reactivity and freeze-thaw deterioration related to poor entr...

Evaluation of concrete pavements with materials-related distress : appendix C.

Science.gov (United States)

2010-03-02

An evaluation of cores sampled from six concrete pavements was performed. Factors contributing to pavement distress observed in the field were determined, including expansive alkali-silica reactivity and freeze-thaw deterioration related to poor entr...

Assessment of reactivity devices for CANDU-6 with DUPIC fuel

International Nuclear Information System (INIS)

Jeong, Chang Joon; Choi, Hang Bok

1998-01-01

Reactivity device characteristics for a CANDU-6 reactor loaded with DUPIC fuel have been assessed. A transport code WIMS-AECL and a three-dimensional diffusion code RFSP were used for the lattice parameter generation and the core calculation, respectively. Three major reactivity devices have been assessed for their inherent functions. For the zone controller system, damping capability for spatial oscillation was investigated. The restart capability of the adjuster system was investigated. The shim operation and power stepback calculation were also performed to confirm the compatibility of the current adjuster rod system. The mechanical control absorber was assessed for the capability to compensate the temperature reactivity feedback following a power reduction. This study has shown that the current reactivity device systems retain their functions when used in a DUPIC fuel CANDU reactor

Neutronics analysis of the TRIGA Mark II reactor core and its experimental facilities

International Nuclear Information System (INIS)

Khan, R.

2010-01-01

The neutronics analysis of the current core of the TRIGA Mark II research reactor is performed at the Atominstitute (ATI) of Vienna University of Technology. The current core is a completely mixed core having three different types of fuels i.e. aluminium clad 20 % enriched, stainless steel clad 20 % enriched and SS clad 70 % enriched (FLIP) Fuel Elements (FE(s)). The completely mixed nature and complicated irradiation history of the core makes the reactor physics calculations challenging. This PhD neutronics research is performed by employing the combination of two best and well practiced reactor simulation tools i.e. MCNP (general Monte Carlo N-particle transport code) for static analysis and ORIGEN2 (Oak Ridge Isotop Generation and depletion code) for dynamic analysis of the reactor core. The PhD work is started to develop a MCNP model of the first core configuration (March 1962) employing fresh fuel composition. The neutrons reaction data libraries ENDF/B-VI is applied taking the missing isotope of Samarium from JEFF3.1. The MCNP model of the very first core has been confirmed by three different local experiments performed on the first core configuration. These experiments include the first criticality, reactivity distribution and the neutron flux density distribution experiment. The first criticality experiment verifies the MCNP model that core achieves its criticality on addition of the 57th FE with a reactivity difference of about 9.3 cents. The measured reactivity worths of four FE(s) and a graphite element are taken from the log book and compared with MCNP simulated results. The percent difference between calculations and measurements ranges from 4 to 22 %. The neutron flux density mapping experiment confirms the model completely exhibiting good agreement between simulated and the experimental results. Since its first criticality, some additional 104-type and 110-type (FLIP) FE(s) have been added to keep the reactor into operation. This turns the current

Application of noise analysis technique for monitoring the moderator temperature coefficient of reactivity in pressurized water reactors

International Nuclear Information System (INIS)

Shieh, D.J.; Upadhyaya, B.R.; Sweeney, F.J.

1987-01-01

A new technique, based on the noise analysis of neutron detector and core-exit coolant temperature signals, is developed for monitoring the moderator temperature coefficient of reactivity in pressurized water reactors (PWRs). A detailed multinodal model is developed and evaluated for the reactor core subsystem of the loss-of-fluid test (LOFT) reactor. This model is used to study the effect of changing the sign of the moderator temperature coefficient of reactivity on the low-frequency phase angle relationship between the neutron detector and the core-exit temperature noise signals. Results show that the phase angle near zero frequency approaches - 180 deg for negative coefficients and 0 deg for positive coefficients when the perturbation source for the noise signals is core coolant flow, inlet coolant temperature, or random heat transfer

Reference core design Mark-III of the experimental multi-purpose, high-temperature, gas-cooled reactor

International Nuclear Information System (INIS)

Shindo, Ryuiti; Watanabe, Takashi; Ishiguro, Okikazu; Kuroki, Syuzi

1977-10-01

The reactivity control system is one of the important items in reactor design, but it is much restricted by structural design of fuel element and pressure vessel in the experimental multi-purpose, high-temperature reactor. Preceding the first conceptual design of the reactor, therefore, the reactivity control system composed of control rod, burnable poison and reserve shutdown system in Mark-II design was re-studied, and several improvements were indicated. (1) The diameter of control rods must be as large as possible because it is impossible to increase the number of control rods. (2) The accuracy in estimation of the reactivity to be compensated with control rods is important because of the mutual interference of pair control rods with the twin configuration in a fuel element. (3) The improvement of core performance in burnup is accompanied by the reduction of design margin for control rods. (4) Increase of the reactivity to be compensated with the burnable poison leads to increase of the core reactivity recovery with burnup, and the assertion of the decrease for recovery of reactivity leads to increase of the temperature dependency of reactivity compensated with control rods. (5) Reduction of reactivity to be compensated with control rods is thus limited by cancellation of the effects in the reactivity recovery and the reactivity temperature dependency. (6) The reserve shutdown system can be designed with margin under the condition of excluding the reactivity of burnup from that to be compensated. (auth.)

Considerations on the influence of fission products in whole core accidents

International Nuclear Information System (INIS)

Meyer Heine, A.; Pattoret, A.; Schmitz, F.

1977-01-01

If the hypothetical Whole Core Accidents which are taken into account in reactor safety analysis can change from one country to another, there is nevertheless a general agreement over their description and main phases. Furthermore the important parameters have also been identified by every laboratory. During the development of such core accidents the role of the fission products in essential. It is not the purpose of this paper to give an exhaustive description of the phases which can be influenced by the fission products, we will try however to focus this study on the most important ones. In a second step we will discuss the equation of state of irradiated fuels; here again one principal preoccupation being to quantify the influence of fission products on reactor accidents. It is not our purpose to enter on the fundamental aspects of the equation of state. The studies and the experimental program launched at the CEA will then be described. Special attention will be directed towards the eventual role of fission products in molten fuel-coolant interactions (MFCls) or the events leading to the initiation of whole core accidents. This paper will be limited to oxide fuels. Whether the whole core accident is initiated by a reactivity defect or a coolant coast-down, one has to deal with four great categories of phenomena. Loss of flow: the power is around the nominal value, while the coolant flow has been reduced by a factor of 5 to 10. This induces boiling and clad weakening. Will the plenum pressure lead to a clad rupture? In case of a rupture, what will be the effect on the voiding of the channel? Transient over power: influence of gases from gaseous and volatile fission products on the fuel movements? MFCIs: Influence of the fission products in the mode of contact between fuel and coolant? Influence on the fuel characteristics. Sodium vapour bubble expansion: influence of the fission products on the heat transfer and eventual condensation of the bubble?

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

Core design methods for advanced LMFBRs

International Nuclear Information System (INIS)

Chandler, J.C.; Marr, D.R.; McCurry, D.C.; Cantley, D.A.

1977-05-01

The multidiscipline approach to advanced LMFBR core design requires an iterative design procedure to obtain a closely-coupled design. HEDL's philosophy requires that the designs should be coupled to the extent that the design limiting fuel pin, the design limiting duct and the core reactivity lifetime should all be equal and should equal the fuel residence time. The design procedure consists of an iterative loop involving three stages of the design sequence. Stage 1 consists of general mechanical design and reactor physics scoping calculations to arrive at an initial core layout. Stage 2 consists of detailed reactor physics calculations for the core configuration arrived at in Stage 1. Based upon the detailed reactor physics results, a decision is made either to alter the design (Stage 1) or go to Stage 3. Stage 3 consists of core orificing and detailed component mechanical design calculations. At this point, an assessment is made regarding design adequacy. If the design is inadequate the entire procedure is repeated until the design is acceptable

Safety characteristics of the US advanced liquid metal reactor core

International Nuclear Information System (INIS)

Magee, P.M.; Dubberley, A.E.; Gyorey, G.L.; Lipps, A.J.; Wu, T.

1991-01-01

The U.S. Advanced Liquid Metal Reactor (ALMR) design employs innovative, passive features to provide an unprecedented level of public safety and the ability to demonstrate this safety to the public. The key features employed in the core design to produce the desired passive safety characteristics are: a small core with a tight restraint system, the use of metallic U-Pu-Zr fuel, control rod withdrawal limiters, and gas expansion modules. In addition, the reactor vessel and closure are designed to have the capability to withstand, with large margins, the maximum possible core disruptive accident without breach and radiological release. (author)

Report on the meeting for examining replacing core

International Nuclear Information System (INIS)

1977-01-01

At the time of examining the application for approval of reactor installation, it must be confirmed that the safety of the concerned reactor is secured with not only the initially loaded core but also the replacing core. Besides, it must be confirmed again that the various criteria concerning the safety are satisfied after the start of operation, because a part of the parameters of the replacing core is dependent on the operational history. On the above described viewpoints, the main parameters affecting the safety and the nuclear and thermal limits of replacing core were reviewed. Moreover, the contents of description concerning replacing core in the application form were examined. As the general matters concerning the safety of replacing core, the scram reactivity curves for BWRs and PWRs, the method of description in the application form concerning the fuel containing gadolinia, and the use of burnable poison in replacing core were examined. The meeting for examining replacing core was organized on September 20, 1976, at the Committee for Examining Reactor Safety, and this report was compiled as the results of 10 meetings. (Kako, I.)

Neutronics calculation of RTP core

Science.gov (United States)

Rabir, Mohamad Hairie B.; Zin, Muhammad Rawi B. Mohamed; Karim, Julia Bt. Abdul; Bayar, Abi Muttaqin B. Jalal; Usang, Mark Dennis Anak; Mustafa, Muhammad Khairul Ariff B.; Hamzah, Na'im Syauqi B.; Said, Norfarizan Bt. Mohd; Jalil, Muhammad Husamuddin B.

2017-01-01

Reactor calculation and simulation are significantly important to ensure safety and better utilization of a research reactor. The Malaysian's PUSPATI TRIGA Reactor (RTP) achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. Since early 90s, neutronics modelling were used as part of its routine in-core fuel management activities. The are several computer codes have been used in RTP since then, based on 1D neutron diffusion, 2D neutron diffusion and 3D Monte Carlo neutron transport method. This paper describes current progress and overview on neutronics modelling development in RTP. Several important parameters were analysed such as keff, reactivity, neutron flux, power distribution and fission product build-up for the latest core configuration. The developed core neutronics model was validated by means of comparison with experimental and measurement data. Along with the RTP core model, the calculation procedure also developed to establish better prediction capability of RTP's behaviour.

Benchmark for Neutronic Analysis of Sodium-cooled Fast Reactor Cores with Various Fuel Types and Core Sizes

International Nuclear Information System (INIS)

Stauff, N.E.; Kim, T.K.; Taiwo, T.A.; Buiron, L.; Rimpault, G.; Brun, E.; Lee, Y.K.; Pataki, I.; Kereszturi, A.; Tota, A.; Parisi, C.; Fridman, E.; Guilliard, N.; Kugo, T.; Sugino, K.; Uematsu, M.M.; Ponomarev, A.; Messaoudi, N.; Lin Tan, R.; Kozlowski, T.; Bernnat, W.; Blanchet, D.; Brun, E.; Buiron, L.; Fridman, E.; Guilliard, N.; Kereszturi, A.; Kim, T.K.; Kozlowski, T.; Kugo, T.; Lee, Y.K.; Lin Tan, R.; Messaoudi, N.; Parisi, C.; Pataki, I.; Ponomarev, A.; Rimpault, G.; Stauff, N.E.; Sugino, K.; Taiwo, T.A.; Tota, A.; Uematsu, M.M.; Monti, S.; Yamaji, A.; Nakahara, Y.; Gulliford, J.

2016-01-01

One of the foremost Generation IV International Forum (GIF) objectives is to design nuclear reactor cores that can passively avoid damage of the reactor when control rods fail to scram in response to postulated accident initiators (e.g. inadvertent reactivity insertion or loss of coolant flow). The analysis of such unprotected transients depends primarily on the physical properties of the fuel and the reactivity feedback coefficients of the core. Within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS), the Sodium Fast Reactor core Feed-back and Transient response (SFR-FT) Task Force was proposed to evaluate core performance characteristics of several Generation IV Sodium-cooled Fast Reactor (SFR) concepts. A set of four numerical benchmark cases was initially developed with different core sizes and fuel types in order to perform neutronic characterisation, evaluation of the feedback coefficients and transient calculations. Two 'large' SFR core designs were proposed by CEA: those generate 3 600 MW(th) and employ oxide and carbide fuel technologies. Two 'medium' SFR core designs proposed by ANL complete the set. These medium SFR cores generate 1 000 MW(th) and employ oxide and metallic fuel technologies. The present report summarises the results obtained by the WPRS for the neutronic characterisation benchmark exercise proposed. The benchmark definition is detailed in Chapter 2. Eleven institutions contributed to this benchmark: Argonne National Laboratory (ANL), Commissariat a l'energie atomique et aux energies alternatives (CEA of Cadarache), Commissariat a l'energie atomique et aux energies alternatives (CEA of Saclay), Centre for Energy Research (CER-EK), Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), Helmholtz Zentrum Dresden Rossendorf (HZDR), Institute of Nuclear Technology and Energy Systems (IKE), Japan Atomic Energy Agency (JAEA), Karlsruhe Institute of Technology (KIT

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.

Frictional and hydraulic behaviour of carbonate fault gouge during fault reactivation - An experimental study

Science.gov (United States)

Delle Piane, Claudio; Giwelli, Ausama; Clennell, M. Ben; Esteban, Lionel; Nogueira Kiewiet, Melissa Cristina D.; Kiewiet, Leigh; Kager, Shane; Raimon, John

2016-10-01

We present a novel experimental approach devised to test the hydro-mechanical behaviour of different structural elements of carbonate fault rocks during experimental re-activation. Experimentally faulted core plugs were subject to triaxial tests under water saturated conditions simulating depletion processes in reservoirs. Different fault zone structural elements were created by shearing initially intact travertine blocks (nominal size: 240 × 110 × 150 mm) to a maximum displacement of 20 and 120 mm under different normal stresses. Meso-and microstructural features of these sample and the thickness to displacement ratio characteristics of their deformation zones allowed to classify them as experimentally created damage zones (displacement of 20 mm) and fault cores (displacement of 120 mm). Following direct shear testing, cylindrical plugs with diameter of 38 mm were drilled across the slip surface to be re-activated in a conventional triaxial configuration monitoring the permeability and frictional behaviour of the samples as a function of applied stress. All re-activation experiments on faulted plugs showed consistent frictional response consisting of an initial fast hardening followed by apparent yield up to a friction coefficient of approximately 0.6 attained at around 2 mm of displacement. Permeability in the re-activation experiments shows exponential decay with increasing mean effective stress. The rate of permeability decline with mean effective stress is higher in the fault core plugs than in the simulated damage zone ones. It can be concluded that the presence of gouge in un-cemented carbonate faults results in their sealing character and that leakage cannot be achieved by renewed movement on the fault plane alone, at least not within the range of slip measureable with our apparatus (i.e. approximately 7 mm of cumulative displacement). Additionally, it is shown that under sub seismic slip rates re-activated carbonate faults remain strong and no frictional

IAEA Technical Meeting on Innovative Fast Reactor Designs with Enhanced Negative Reactivity Feedback Features. Working Material