Physical and geometrical parameters of ANNA critical assemblies. Pt. 2

International Nuclear Information System (INIS)

Malewski, S.; Dabrowski, C.

1973-01-01

An extended analysis of four critical configurations of ANNA Assembly has been performed. Diffusion parameters of the thermal group and of one or three epithermal groups have been determined. Using these data the critical calculations have been carried out and the main neutron density distributions presented. The role of some neutron processes in these systems and their influence on integral parameters has been considered. The calculated quantities have been compared with the available experimental data. (author)

Reactor Dynamics Experiments with a Sub-Critical Assembly

International Nuclear Information System (INIS)

Miley, G.H.; Yang, Y.; Wu, L.; Momota, H.

2004-01-01

A resurgence in use of nuclear power is now underway worldwide. However due to the shutdown of many university research reactors , student laboratories must rely more heavily on use of sub-critical assemblies. Here a driven sub-critical is described that uses a cylindrical Inertial Electrostatic Confinement (IEC) device to provide a fusion neutron source. The small IEC neutron source would be inserted in a fuel element position, with its power input controlled externally at a control panel. This feature opens the way to use of the critical assembly for a number of transient experiments such as sub-critical pulsing and neutron wave propagation. That in turn adds important new insights and excitement for the student teaching laboratory

Trial evaluation on criticality safety of the fuel assemblies at falling accident as spent fuel transport and storage cask

International Nuclear Information System (INIS)

Tadano, Tomoaki

2016-01-01

The authors conducted critical safety assessment on the supposed event at the time of a fall accident of cask, and examined the influence on criticality safety. If the spacer of fuel assembly is sound, it is assumed that the pitch of fuel rod interval changes, and if the spacer is broken, it is assumed that the fuel rod is unevenly distributed in the basket. For the critical calculation of fuel assembly basket system, they performed it using a calculation code. For both of the single cell and assembly, calculation results showed an increase in the effective multiplication factor of reactivity of 2-3%. When this reactivity is applied to the criticality analysis result of PWR fuel assembly, the value approaches to the limit 0.95 of the neutron effective multiplication factor keff. However, the keff when new fuel is loaded is sufficiently lower than 0.93. Therefore, it is unlikely that the criticality analysis result approaches to 0.95 at all burnups, and the possibility to become criticality is very low in actual spent fuel transport. When considering the reactivity of this research, it is possible that the design condition for the assumption of novel fuel loading becomes severer. Furthermore, criticality analysis under non-uniform pitch will become necessary, and criticality safety analysis for BWR fuel with heterogeneous enrichment degree and burnup degree will become also necessary. (A.O.)

Coretran/Vipre assembly critical power assessment against Nupec BWR experiments

International Nuclear Information System (INIS)

Aounallah, Y.

2001-01-01

This study has been performed, in the framework of the STARS project, to assess CORETRAN-01/VIPRE-02 code capability to predict critical heat flux conditions for BWR fuel assemblies. The assessment is based on comparisons of the code results with the NUPEC steady-state critical power measurements on full-scale assemblies tested under a range of flow conditions. Two assembly types were considered, the standard BWR 8 x 8 and the so-called ''high-burnup'' assembly, similar to GE-10. Code modelling options that have a significant impact on the results have been identified, along with code limitations. (author)

Sensitivity coefficients of reactor parameters in fast critical assemblies and uncertainty analysis

International Nuclear Information System (INIS)

Aoyama, Takafumi; Suzuki, Takayuki; Takeda, Toshikazu; Hasegawa, Akira; Kikuchi, Yasuyuki.

1986-02-01

Sensitivity coefficients of reactor parameters in several fast critical assemblies to various cross sections were calculated in 16 group by means of SAGEP code based on the generalized perturbation theory. The sensitivity coefficients were tabulated and the difference of sensitivity coefficients was discussed. Furthermore, the uncertainty of calculated reactor parameters due to cross section uncertainty were estimated using the sensitivity coefficients and cross section covariance data. (author)

Coretran/Vipre assembly critical power assessment against Nupec BWR experiments

Energy Technology Data Exchange (ETDEWEB)

Aounallah, Y. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

2001-07-01

This study has been performed, in the framework of the STARS project, to assess CORETRAN-01/VIPRE-02 code capability to predict critical heat flux conditions for BWR fuel assemblies. The assessment is based on comparisons of the code results with the NUPEC steady-state critical power measurements on full-scale assemblies tested under a range of flow conditions. Two assembly types were considered, the standard BWR 8 x 8 and the so-called ''high-burnup'' assembly, similar to GE-10. Code modelling options that have a significant impact on the results have been identified, along with code limitations. (author)

The Bare Critical Assembly of Natural Uranium and Heavy Water

Energy Technology Data Exchange (ETDEWEB)

Popovic, D [Boris Kidric Institute of Nuclear Sciences, Vinca, Belgrade (Yugoslavia)

1958-07-01

The first reactor built in Yugoslavia was the bare zero energy heavy water and natural uranium assembly at the Boris Kidric Institute of Nuclear Sciences, Belgrade. The reactor went critical on April 29, 1958. The possession of four tons of natural uranium metal and the temporary availability of seven tons of heavy water encouraged the staff of the Institute to build a critical assembly. A critical assembly was chosen, rather than high flux reactor, because the heavy water was available only temporarily. Besides, a 10 MW, enriched uranium, research reactor is being built at the same Institute and should be ready for operation late this year. It was supposed that the zero energy reactor would provide experience in carrying out critical experiments, operational experience with nuclear reactors, and the possibility for an extensive program in reactor physics. (author)

Organization and methods of radiation monitoring while working at nuclear critical assemblies

International Nuclear Information System (INIS)

Shishkin, G.V.; Komissarov, L.A.

1980-01-01

The organization and methods of environmental radiation monitoring while working at nuclear critical assemblies, are described. Necessary equipment for critical assemblies (signal and Ventilation systems, devices for recording accidental radiation levels of and for measuring radiation field distribution) and the personnel program of actions in case of nuclear accident. The dosimetric control at critical assemblies is usually ensured by telesystems. 8004-01 multi-channel dosimetric device is described as an example of such-system [ru

Basic experiments of reactor physics using the critical assembly TCA

International Nuclear Information System (INIS)

Obara, Toru; Igashira, Masayuki; Sekimoto, Hiroshi; Nakajima, Ken; Suzaki, Takenori.

1994-02-01

This report is based on lectures given to graduate students of Tokyo Institute of Technology. It covers educational experiments conducted with the Tank-Type Critical Assembly (TCA) at Japan Atomic Energy Research Institute in July, 1993. During this period, the following basic experiments on reactor physics were performed: (1) Critical approach experiment, (2) Measurement of neutron flux distribution, (3) Measurement of power distribution, (4) Measurement of fuel rod worth distribution, (5) Measurement of safety sheet worth by the rod drop method. The principle of experiments, experimental procedure, and analysis of results are described in this report. (author)

Safety considerations of new critical assembly for the Research Reactor Institute, Kyoto University

International Nuclear Information System (INIS)

Umeda, Iwao; Matsuoka, Naomi; Harada, Yoshihiko; Miyamoto, Keiji; Kanazawa, Takashi

1975-01-01

The new critical assembly type of nuclear reactor having three cores for the first time in the world was completed successfully at the Research Reactor Institute of Kyoto University in autumn of 1974. It is called KUCA (Kyoto University Critical Assembly). Safety of the critical assembly was considered sufficiently in consequence of discussions between the researchers of the institute and the design group of our company, and then many bright ideas were created through the discussions. This paper is described the new safety design of main equipments - oil pressure type center core drive mechanism, removable water overflow mechanism, core division mechanism, control rod drive mechansim, protection instrumentation system and interlock key system - for the critical assembly. (author)

Research and design calculation of multipurpose critical assembly using moderated light water and low enriched fuel from 1.6 to 5.0% U-235

International Nuclear Information System (INIS)

Nguyen Kien Cuong; Vo Doan Hai Dang; Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Minh Tuan; Nguyen Manh Hung; Pham Quang Huy; Tran Quoc Duong; Tran Tri Vien

2015-01-01

Basing on the idea in ??using fuel of nuclear power plants such as PWR (AP-1000) and VVER-1000 with light water as moderation, design calculation of critical assembly was performed to confirm the possibility of using these fuels. Designed critical assembly has simple structure consisting of low enriched fuel from 1.6% to 5% U-235; water has functions as cooling, biological protection and control. Critical assembly is operated at nominal power 100 W with fuel pitch about 2.0 cm. Applications of the critical assembly are quite abundant in basic research, education and training with low investment cost compare with research reactor and easy in operation. So critical assembly can be used for university or training centre for nuclear engineering training. Main objectives of the project are: design calculation in neutronics, thermal hydraulics and safety analysis for critical configuration benchmarks using low enriched fuel; design in mechanical and auxiliary systems for critical assembly; determine technical specifications and estimate construction, installation cost of critical assembly. The process of design, fabrication, installation and construction of critical assembly will be considered with different implementation phases and localization capabilities in installation of critical assembly is highly feasibility. Cost estimation of construction and installation of critical assembly was implemented and showed that investment cost for critical assembly is much lower than research reactor and most of components, systems of critical assembly can be localized with current technique quality of the country. (author)

Verification of homogenization in fast critical assembly analyses

International Nuclear Information System (INIS)

Chiba, Go

2006-01-01

In the present paper, homogenization procedures for fast critical assembly analyses are investigated. Errors caused by homogenizations are evaluated by the exact perturbation theory. In order to obtain reference solutions, three-dimensional plate-wise transport calculations are performed. It is found that the angular neutron flux along plate boundaries has a significant peak in the fission source energy range. To treat this angular dependence accurately, the double-Gaussian Chebyshev angular quadrature set with S 24 is applied. It is shown that the difference between the heterogeneous leakage theory and the homogeneous theory is negligible, and that transport cross sections homogenized with neutron flux significantly underestimate neutron leakage. The error in criticality caused by a homogenization is estimated at about 0.1%Δk/kk' in a small fast critical assembly. In addition, the neutron leakage is overestimated by both leakage theories when sodium plates in fuel lattices are voided. (author)

Reactor physics studies in the GCFR phase-II critical assembly

International Nuclear Information System (INIS)

Pond, R.B.

1976-09-01

The reactor physics studies performed in the gas cooled fast reactor (GCFR) mockup on ZPR-9 are covered. This critical assembly, designated Phase II in the GCFR program, had a single zone PuO 2 -UO 2 core composition and UO 2 radial and axial blankets. The assembly was built both with and without radial and axial stainless steel reflectors. The program included the following measurements: small-sample reactivity worths of reactor constituent materials (including helium); 238 U Doppler effect; uranium and plutonium reaction rate distributions; thorium, uranium, and plutonium α and reactor kinetics. Analysis of the measurements used ENDF/B-IV nuclear data; anisotropic diffusion coefficients were used to account for neutron streaming effects. Comparison of measurements and calculations to GCFR Phase I are also made

Operating procedures for the Pajarito Site Critical Assembly Facility

International Nuclear Information System (INIS)

Malenfant, R.E.

1983-03-01

Operating procedures consistent with DOE Order 5480.2, Chapter VI, and the American National Standard Safety Guide for the Performance of Critical Experiments are defined for the Pajarito Site Critical Assembly Facility of the Los Alamos National Laboratory. These operating procedures supersede and update those previously published in 1973 and apply to any criticality experiment performed at the facility

Thor, a thorium-reflected plutonium-metal critical assembly

International Nuclear Information System (INIS)

Hansen, G.E.; Paxton, H.C.

1979-01-01

Critical specifications of Thor, an old assembly of thorium-reflected plutonium, have been refined. These specifications are brought together with void coefficients, Rossi-alpha values, fission traverses, and spectral indices

Calculation and analysis for a series of enriched uranium bare sphere critical assemblies

International Nuclear Information System (INIS)

Yang Shunhai

1994-12-01

The imported reactor fuel assembly MARIA program system is adapted to CYBER 825 computer in China Institute of Atomic Energy, and extensively used for a series of enriched uranium bare sphere critical assemblies. The MARIA auxiliary program of resonance modification MA is designed for taking account of the effects of resonance fission and absorption on calculated results. By which, the multigroup constants in the library attached to MARIA program are revised based on the U.S. Evaluated Nuclear Data File ENDF/B-IV, the related nuclear data files are replaced. And then, the reactor geometry buckling and multiplication factor are given in output tapes. The accuracy of calculated results is comparable with those of Monte Carlo and Sn method, and the agreement with experiment result is in 1%. (5 refs., 4 figs., 3 tabs.)

Refinement of criticality and breeding parameters by means of experiments on a series of critical assemblies

International Nuclear Information System (INIS)

Golubev, V.I.; Dulin, V.A.; Kazanskij, Yu.A.; Mamontov, V.M.; Mozhaev, V.K.; Sidorov, G.I.

1980-01-01

A programme of measurements was performed on a number of critical assemblies with the aim of obtaining reliable experimental data under conditions approximating the simplest calculation model. To this end the neutron balance at the centres of the BFS-31, BFS-33, BFS-35, BFS-38, KBR-3 and KBR-7 critical assemblies was investigated. These assemblies contained central inserts made of uranium dioxide (BFS-33), natural uranium oxide and plutonium metal (BFS-31), natural uranium and plutonium metal (BFS-38), 90% enriched metallic uranium and stainless steel (KBR-3) and enriched uranium dioxide and nickel (KBR-7). The composition of the inserts was such that Ksub(infinite)=1. The K + values, the ratios of the reaction rates of the principal raw material and fissionable isotopes and the reactivity coefficients of a number of materials were measured in the inserts. The components of the breeding coefficient were measured at the centre of the BFS-39 critical assembly which simulates a power reactor (simplest composition with low- and high-enrichment zones and no control mechanism). The authors describe briefly the critical assemblies, the methods of measurement and calculation and methods of correcting for differences between the calculation model and the conditions under which the measurements were performed and compare the results of the experiments with the corresponding theoretical values obtained using various systems of group constants. In their latest versions, the group constants derived from different sets of integral experiments describe the experimental data much better than was previously possible. The deviations which occur in the predicted criticality and breeding parameters using different versions of the constants essentially reflect the difference in the results of the sets of integral experiments that were used for the group constants. (author)

Development of training simulator based on critical assemblies test bench

International Nuclear Information System (INIS)

Narozhnyi, A.T.; Vorontsov, S.V.; Golubeva, O.A.; Dyudyaev, A.M.; Il'in, V.I.; Kuvshinov, M.I.; Panin, A.V.; Peshekhonov, D.P.

2007-01-01

When preparing critical mass experiment, multiplying system (MS) parts are assembled manually. This work is connected with maximum professional risk to personnel. Personnel training and keeping the skill of working experts is the important factor of nuclear safety maintenance. For this purpose authors develop a training simulator based on functioning critical assemblies test bench (CATB), allowing simulation of the MS assemblage using training mockups made of inert materials. The control program traces the current status of MS under simulation. A change in the assembly neutron physical parameters is mapped in readings of the regular devices. The simulator information support is provided by the computer database on physical characteristics of typical MS components The work in the training mode ensures complete simulation of real MS assemblage on the critical test bench. It makes it possible to elaborate the procedures related to CATB operation in a standard mode safely and effectively and simulate possible abnormal situations. (author)

Reactor physics studies in the steam flooded GCFR-Phase II critical assembly

International Nuclear Information System (INIS)

Bhattacharyya, S.K.

1978-08-01

A possible accident scenario in a Gas-Cooled Fast Reactor (GCFR) is the leakage of secondary steam into the core. Considerable analytical effort has gone into the study of the effects of such an accidental steam entry. The work described represents the first full scale experimental study of the steam-entry phenomenon in GCFRs. The reference GCFR model used for the study was the benchmark GCFR Phase II assembly, and polyethylene foam was used to provide a very homogeneous steam simulation. The reactivity worth of steam entry was measured for three different steam densities. In addition, a set of integral physics parameters were measured in the largest steam density (0.008 g/cm 3 ) configuration. The corresponding parameters were also measured in dry reference GCFR critical assembly for comparison. The experiments were analyzed using ENDF/B-IV data and two-dimensional diffusion theory methods. As in earlier GCFR critical experiments analysis, the Benoist method was used to treat the problem of neutron streaming

New calculations for critical assemblies using MCNP4B

International Nuclear Information System (INIS)

Adams, A.A.; Frankle, S.C.; Little, R.C.

1997-07-01

A suite of 41 criticality benchmarks has been modeled using MCNP trademark (version 4B). Most of the assembly specifications were obtained from the Cross Section Evaluation Working Group (CSEWG) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) compendiums of experimental benchmarks. A few assembly specifications were obtained from experimental papers. The suite contains thermal and fast assemblies, bare and reflected assemblies, and emphasizes 233 U, 235 U, 238 U, and 239 Pu. The values of k eff for each assembly in the suite were calculated using MCNP libraries derived primarily from release 2 of ENDF/B-V and release 2 of ENDF/B-VI. The results show that the new ENDF/B-VI.2 evaluations for H, O, N, B, 235 U, 238 U, and 239 Pu can have a significant impact on the values of k eff . In addition to the integral quantity k eff , several additional experimental measurements were performed and documented. These experimental measurements include central fission and reaction-rate ratios for various isotopes, and neutron leakage and flux spectra. They provide more detailed information about the accuracy of the nuclear data than can k eff . Comparison calculations were performed using both ENDF/B-V.2 and ENDF/B-VI.2-based data libraries. The purpose of this paper is to compare the results of these additional calculations with experimental data, and to use these results to assess the quality of the nuclear data

Safe Operation of Critical Assemblies and Research Reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1961-05-15

This Manual is provided as a guide to the safe operation of critical assemblies and small research reactors. It is intended that it should be used by all authorities and persons concerned with, or responsible for, the use of such equipment, in addition to the scientists and technologists who are actually working with, or operating it. It is suggested that it will be of use to those wishing to design and manufacture, or purchase, critical assemblies or research reactors, as well as those already in possession of them, and that it will prove particularly helpful to those users who have no direct access to other collected sources of information. This Manual is not a set of rules or a code of practice, but a series of recommendations which must be interpreted with scientific judgement in their application to any particular problem. The guiding principles are given from which good operational procedures may be established and improved. The promulgation of rigid standards is both impossible and undesirable at the present time, since the topics discussed form part of a rapidly growing science and technology. Therefore, any recommendations made should not be used to restrict or inhibit future developments. The Manual is intended mainly for use in those Member States where there has been little experience in the operation of critical assemblies and research reactors. It has been compounded from the best practices which exist in Member States having a large amount of such experience, so that nothing in it should conflict with the best practices to be encountered in the field of safe operation.

Safe Operation of Critical Assemblies and Research Reactors

International Nuclear Information System (INIS)

1961-01-01

This Manual is provided as a guide to the safe operation of critical assemblies and small research reactors. It is intended that it should be used by all authorities and persons concerned with, or responsible for, the use of such equipment, in addition to the scientists and technologists who are actually working with, or operating it. It is suggested that it will be of use to those wishing to design and manufacture, or purchase, critical assemblies or research reactors, as well as those already in possession of them, and that it will prove particularly helpful to those users who have no direct access to other collected sources of information. This Manual is not a set of rules or a code of practice, but a series of recommendations which must be interpreted with scientific judgement in their application to any particular problem. The guiding principles are given from which good operational procedures may be established and improved. The promulgation of rigid standards is both impossible and undesirable at the present time, since the topics discussed form part of a rapidly growing science and technology. Therefore, any recommendations made should not be used to restrict or inhibit future developments. The Manual is intended mainly for use in those Member States where there has been little experience in the operation of critical assemblies and research reactors. It has been compounded from the best practices which exist in Member States having a large amount of such experience, so that nothing in it should conflict with the best practices to be encountered in the field of safe operation.

Application of SN and Monte Carlo codes to the SHEBA critical assemblies

International Nuclear Information System (INIS)

O'Dell, R.D.

1993-01-01

The Solution High-Energy Burst Assembly (SHEBA) at Los Alamos is a low-enriched (4.95 wt. %) aqueous uranyl fluoride solution critical assembly. There are two SHEBA configurations, both consisting of right circular cylinders with a central control rod. The first configuration, hereafter called the old SHEBA, had a fuel solution diameter of 54.6 cm and a measured critical solution height of 36.5 cm. An improved modification, hereafter called the new SHEBA, has a fuel solution diameter of 48.9 cm but since it is not yet operational, the critical solution height has not yet been measured. In this presentation the application of the discrete-ordinates (S N ) code TWODANT using Hansen-Roach cross sections and the MCNP Monte Carlo code using continuous-energy cross sections for calculating the critical solution heights for both the old and new SHEBA assemblies is described. The code's predictions are compared and it is shown that a single calculation with a standard computer code may yield misleading results, especially when using a Monte Carlo code

Accident Analysis of High Density Storage Rack for Fresh Fuel Assemblies

International Nuclear Information System (INIS)

Jang, K. J.; Lee, M. J.; Jin, H. U.; Park, J. H.; Shin, S. Y.

2009-01-01

Recently KONES and KNF have developed the so called suspension-type High Density Storage Rack (HDSR) for fresh fuel assemblies. The USNRC OT position paper specifies that the design of the rack must ensure the functional integrity of the fuel racks under all credible fuel assembly drop events. In this context the functional integrity means the criticality safety. That is to say, the drop events must not bring any danger to the criticality safety of HDSR. This paper shows the results of the analysis carried out to demonstrate the regulatory compliance of the proposed racks under postulated accidental drop events

An improved benchmark model for the Big Ten critical assembly - 021

International Nuclear Information System (INIS)

Mosteller, R.D.

2010-01-01

A new benchmark specification is developed for the BIG TEN uranium critical assembly. The assembly has a fast spectrum, and its core contains approximately 10 wt.% enriched uranium. Detailed specifications for the benchmark are provided, and results from the MCNP5 Monte Carlo code using a variety of nuclear-data libraries are given for this benchmark and two others. (authors)

Sensitivity Analysis of Criticality for Different Nuclear Fuel Shapes

Energy Technology Data Exchange (ETDEWEB)

Kang, Hyun Sik; Jang, Misuk; Kim, Seoung Rae [NESS, Daejeon (Korea, Republic of)

2016-10-15

Rod-type nuclear fuel was mainly developed in the past, but recent study has been extended to plate-type nuclear fuel. Therefore, this paper reviews the sensitivity of criticality according to different shapes of nuclear fuel types. Criticality analysis was performed using MCNP5. MCNP5 is well-known Monte Carlo codes for criticality analysis and a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical systems. We performed the sensitivity analysis of criticality for different fuel shapes. In sensitivity analysis for simple fuel shapes, the criticality is proportional to the surface area. But for fuel Assembly types, it is not proportional to the surface area. In sensitivity analysis for intervals between plates, the criticality is greater as the interval increases, but if the interval is greater than 8mm, it showed an opposite trend that the criticality decrease by a larger interval. As a result, it has failed to obtain the logical content to be described in common for all cases. The sensitivity analysis of Criticality would be always required whenever subject to be analyzed is changed.

Sensitivity Analysis of Criticality for Different Nuclear Fuel Shapes

International Nuclear Information System (INIS)

Kang, Hyun Sik; Jang, Misuk; Kim, Seoung Rae

2016-01-01

Rod-type nuclear fuel was mainly developed in the past, but recent study has been extended to plate-type nuclear fuel. Therefore, this paper reviews the sensitivity of criticality according to different shapes of nuclear fuel types. Criticality analysis was performed using MCNP5. MCNP5 is well-known Monte Carlo codes for criticality analysis and a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical systems. We performed the sensitivity analysis of criticality for different fuel shapes. In sensitivity analysis for simple fuel shapes, the criticality is proportional to the surface area. But for fuel Assembly types, it is not proportional to the surface area. In sensitivity analysis for intervals between plates, the criticality is greater as the interval increases, but if the interval is greater than 8mm, it showed an opposite trend that the criticality decrease by a larger interval. As a result, it has failed to obtain the logical content to be described in common for all cases. The sensitivity analysis of Criticality would be always required whenever subject to be analyzed is changed

Safety analysis of the Los Alamos critical experiments facility: burst operation of Skua

International Nuclear Information System (INIS)

Orndoff, J.D.; Paxton, H.C.; Wimett, T.F.

1979-05-01

A detailed consideration of the Skua burst assembly is presented, thereby supplementing the facility safety analysis report covering the operation of other critical assemblies at Los Alamos. As with these assemblies the small fission-product inventory, ambient pressure, and moderate temperatures in Skua are amenable to straightforward measures to ensure the protection of the public

Safety analysis of the Los Alamos critical experiments facility: burst operation of Skua

International Nuclear Information System (INIS)

Orndoff, J.D.; Paxton, H.C.; Wimett, T.F.

1980-12-01

Detailed consideration of the Skua burst assembly is provided, thereby supplementing the facility Safety Analysis Report covering the operation of other critical assemblies at the Los Alamos Scientific Laboratory. As with these assemblies the small fission-product inventory, ambient pressure, and moderate temperatures in Skua are amenable to straightforward measures to ensure the protection of the public

Supplementary neutron-flux calculations for the ORNL Pool Critical Assembly Pressure Vessel Facility

International Nuclear Information System (INIS)

Maudlin, P.J.; Maerker, R.E.

1982-01-01

A three-dimensional Monte Carlo calculation using the MORSE code was performed to validate a procedure previously adopted in the ORNL discrete ordinate analysis of measurements made in the ORNL Pool Critical Assembly Pressure Vessel Facility. The results of these flux calculations agree, within statistical undertainties of about 5%, with those obtained from a discrete ordinate analysis employing the same procedure. This study therefore concludes that the procedure for combining several one- and two-dimensional discrete ordinate calculations into a three-dimensional flux is sufficiently accurate that it does not account for the existing discrepancies observed between calculations and measurements in this facility

Supplementary neutron-flux calculations for the ORNL Pool Critical Assembly Pressure Vessel Facility

Energy Technology Data Exchange (ETDEWEB)

Maudlin, P.J.; Maerker, R.E.

1982-01-01

A three-dimensional Monte Carlo calculation using the MORSE code was performed to validate a procedure previously adopted in the ORNL discrete ordinate analysis of measurements made in the ORNL Pool Critical Assembly Pressure Vessel Facility. The results of these flux calculations agree, within statistical undertainties of about 5%, with those obtained from a discrete ordinate analysis employing the same procedure. This study therefore concludes that the procedure for combining several one- and two-dimensional discrete ordinate calculations into a three-dimensional flux is sufficiently accurate that it does not account for the existing discrepancies observed between calculations and measurements in this facility.

Fission reactor critical experiments and analysis

International Nuclear Information System (INIS)

Anon.

1977-01-01

Work accomplished in support of nonweapons programs by LASL Group Q-14 is described. Included are efforts in basic critical measurements, nuclear criticality safety, a plasma core critical assembly, and reactivity coefficient measurements

Pool critical assembly pressure vessel facility benchmark

International Nuclear Information System (INIS)

Remec, I.; Kam, F.B.K.

1997-07-01

This pool critical assembly (PCA) pressure vessel wall facility benchmark (PCA benchmark) is described and analyzed in this report. Analysis of the PCA benchmark can be used for partial fulfillment of the requirements for the qualification of the methodology for pressure vessel neutron fluence calculations, as required by the US Nuclear Regulatory Commission regulatory guide DG-1053. Section 1 of this report describes the PCA benchmark and provides all data necessary for the benchmark analysis. The measured quantities, to be compared with the calculated values, are the equivalent fission fluxes. In Section 2 the analysis of the PCA benchmark is described. Calculations with the computer code DORT, based on the discrete-ordinates method, were performed for three ENDF/B-VI-based multigroup libraries: BUGLE-93, SAILOR-95, and BUGLE-96. An excellent agreement of the calculated (C) and measures (M) equivalent fission fluxes was obtained. The arithmetic average C/M for all the dosimeters (total of 31) was 0.93 ± 0.03 and 0.92 ± 0.03 for the SAILOR-95 and BUGLE-96 libraries, respectively. The average C/M ratio, obtained with the BUGLE-93 library, for the 28 measurements was 0.93 ± 0.03 (the neptunium measurements in the water and air regions were overpredicted and excluded from the average). No systematic decrease in the C/M ratios with increasing distance from the core was observed for any of the libraries used

Effect of a dual-purpose cask payload increment of spent fuel assemblies from VVER 1000 Bushehr Nuclear Power Plant on basket criticality

International Nuclear Information System (INIS)

Rezaeian, M.; Kamali, J.

2017-01-01

Dual-purpose casks can be utilized for dry interim storage and transportation of the highly radioactive spent fuel assemblies (SFAs) of Bushehr Nuclear Power Plant (NPP). Criticality safety analysis was carried out using the MCNP code for the cask containing 12, 18, or 19 SFAs. The basket materials of borated stainless steel and Boral (Al-B_4C) were investigated, and the minimum required receptacle pitch of the basket was determined. - Highlights: • Criticality safety analysis for a dual purpose cask was carried out. • The basket material of borated stainless steel and Boral were investigated. • Minimum receptacle pitch was determined for 12, 18, or 19 VVER 1000 spent fuel assemblies.

Potential impacts of ENDF/B-V on critical experiment analysis based on ZEBRA-8 criticals

Energy Technology Data Exchange (ETDEWEB)

Choong, T S

1982-06-01

The ZEBRA-8 series of null-zone measurements featured a different neutron spectrum for each assembly. The experiments were designed for the purpose of basic data testing. The series cover a range of spectra both harder and softer than that for the LMFBR. The potential impacts of the newly released ENDF/BV cross section library on LMFBR critical exeriment analysis are discussed based on analysis of ZEBRA-8 series.

Influence of “whirlwind” mixing grids on the critical power of WWER fuel assembly

International Nuclear Information System (INIS)

Selivanov, Yu.F.; Pomet'ko, R.S.; Volkov, S.E.

2014-01-01

The problem of optimizing the number and placement of lattices in different types assemblies is discussed. The effect of the amount of mixing lattices and their locations in assemblies on the conditions of occurrence of boiling crisis in the fuel assembly on its critical power (power of assembly in case of boiling crisis) is studied. Experiments were carried out with the use of freon as a coolant. It is recommended simultaneous use in the assembly of lattices of “whirlwind” type, well-intensifying heat exchange, and cell lattices of “pass” type (or lattices with deflectors) affecting on moving flow, provided the optimal location of lattices in the assembly [ru

Disposal criticality analysis methodology for fissile waste forms

International Nuclear Information System (INIS)

Davis, J.W.; Gottlieb, P.

1998-03-01

A general methodology has been developed to evaluate the criticality potential of the wide range of waste forms planned for geologic disposal. The range of waste forms include commercial spent fuel, high level waste, DOE spent fuel (including highly enriched), MOX using weapons grade plutonium, and immobilized plutonium. The disposal of these waste forms will be in a container with sufficiently thick corrosion resistant barriers to prevent water penetration for up to 10,000 years. The criticality control for DOE spent fuel is primarily provided by neutron absorber material incorporated into the basket holding the individual assemblies. For the immobilized plutonium, the neutron absorber material is incorporated into the waste form itself. The disposal criticality analysis methodology includes the analysis of geochemical and physical processes that can breach the waste package and affect the waste forms within. The basic purpose of the methodology is to guide the criticality control features of the waste package design, and to demonstrate that the final design meets the criticality control licensing requirements. The methodology can also be extended to the analysis of criticality consequences (primarily increased radionuclide inventory), which will support the total performance assessment for the respository

Spatial kinetics studies in liquid-metal fast breeder reactor critical assemblies

International Nuclear Information System (INIS)

Brumback, S.B.; Goin, R.W.; Carpenter, S.G.

1988-01-01

Recent measurements in the zero-power physics reactor have been used to study the effect of spatial decoupling in fast reactor critical assemblies of various sizes and compositions. Flux distributions in these assemblies had varying degrees of sensitivity to perturbations. Decoupling was investigated using rod-drop, boron-oscillator, and noise-coherence techniques, which emphasized different times following perturbations. Equilibrium flux distributions were also measured for subcritical configurations with inserted control rods. For most assemblies, accurate reactivity measurements were obtained by analyzing the power history from a single detector using inverse kinetics methods, assuming an instantaneous efficiency change for the detector. The instantaneous efficiency change assumption broke down, however, in assemblies with zones in which normal plutonium fuel was replaced by /sup 235/U fuel or fuel with a high /sup 240/Pu content. Flux redistributions caused by perturbations in these cores took several minutes to evolve

SECOND WASTE PACKAGE PROBABILISTIC CRITICALITY ANALYSIS: GENERATION AND EVALUATION OF INTERNAL CRITICIALITY CONFIGURATIONS

Energy Technology Data Exchange (ETDEWEB)

P. Gottlieb, J.R. Massari, J.K. McCoy

1996-03-27

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development (WPD) department to provide an evaluation of the criticality potential within a waste package having sonic or all of its contents degraded by corrosion and removal of neutron absorbers. This analysis is also intended to provide an estimate of the consequences of any internal criticality, particularly in terms of any increase in radionuclide inventory. These consequence estimates will be used as part of the WPD input to the Total System Performance Assessment. The ultimate objective of this analysis is to augment the information gained from the Initial Waste Package Probabilistic Criticality Analyses (Ref. 5.8 and 5.9, hereafter referred to as IPA) to a degree which will support preliminary waste package design recommendations intended to reduce the risk of waste package criticality and the risk to total repository system performance posed by the consequences of any criticality. The IPA evaluated the criticality potential under the assumption that the waste package basket retained its structural integrity, so that the assemblies retained their initial separation, even when the neutron absorbers had been leached from the basket. This analysis is based on the more realistic condition that removal of the neutron absorbers is a consequence of the corrosion of the steel in which they are contained, which has the additional consequence of reducing the structural support between assemblies. The result is a set of more reactive configurations having a smaller spacing between assemblies, or no inter-assembly spacing at all. Another difference from the IPA is the minimal attention to probabilistic evaluation given in this study. Although the IPA covered a time horizon to 100,000 years, the lack of consideration of basket degradation modes made it primarily applicable to the first 10,000 years. In contrast, this study, by focusing on the degraded modes of the basket, is primarily

Critical experiments analysis by ABBN-90 constant system

Energy Technology Data Exchange (ETDEWEB)

Tsiboulia, A.; Nikolaev, M.N.; Golubev, V. [Institute of Physics and Power Engineering, Obninsk (Russian Federation)] [and others

1997-06-01

The ABBN-90 is a new version of the well-known Russian group-constant system ABBN. Included constants were calculated based on files of evaluated nuclear data from the BROND-2, ENDF/B-VI, and JENDL-3 libraries. The ABBN-90 is intended for the calculation of different types of nuclear reactors and radiation shielding. Calculations of criticality safety and reactivity accidents are also provided by using this constant set. Validation of the ABBN-90 set was made by using a computerized bank of evaluated critical experiments. This bank includes the results of experiments conducted in Russia and abroad of compact spherical assemblies with different reflectors, fast critical assemblies, and fuel/water-solution criticalities. This report presents the results of the calculational analysis of the whole collection of critical experiments. All calculations were produced with the ABBN-90 group-constant system. Revealed discrepancies between experimental and calculational results and their possible reasons are discussed. The codes and archives INDECS system is also described. This system includes three computerized banks: LEMEX, which consists of evaluated experiments and their calculational results; LSENS, which consists of sensitivity coefficients; and LUND, which consists of group-constant covariance matrices. The INDECS system permits us to estimate the accuracy of neutronics calculations. A discussion of the reliability of such estimations is finally presented. 16 figs.

Disposal criticality analysis for aluminum-based DOE fuels

International Nuclear Information System (INIS)

Davis, J.W.; Gottlieb, P.

1997-11-01

This paper describes the disposal criticality analysis for canisters containing aluminum-based Department of Energy fuels from research reactors. Different canisters were designed for disposal of highly enriched uranium (HEU) and medium enriched uranium (MEU) fuel. In addition to the standard criticality concerns in storage and transportation, such as flooding, the disposal criticality analysis must consider the degradation of the fuel and components within the waste package. Massachusetts Institute of Technology (MIT) U-Al fuel with 93.5% enriched uranium and Oak Ridge Research Reactor (ORR) U-Si-Al fuel with 21% enriched uranium are representative of the HEU and MEU fuel inventories, respectively. Conceptual canister designs with 64 MIT assemblies (16/layer, 4 layers) or 40 ORR assemblies (10/layer, 4 layers) were developed for these fuel types. Borated stainless steel plates were incorporated into a stainless steel internal basket structure within a 439 mm OD, 15 mm thick XM-19 canister shell. The Codisposal waste package contains 5 HLW canisters (represented by 5 Defense Waste Processing Facility canisters from the Savannah River Site) with the fuel canister placed in the center. It is concluded that without the presence of a fairly insoluble neutron absorber, the long-term action of infiltrating water can lead to a small, but significant, probability of criticality for both the HEU and MEU fuels. The use of 1.5kg of Gd distributed throughout the MIT fuel and the use of carbon steels for the structural basket or 1.1 kg of Gd distributed in the ORR fuel will reduce the probability of criticality to virtually zero for both fuels

Educational reactor-physics experiments with the critical assemble TCA

Energy Technology Data Exchange (ETDEWEB)

Tsutsui, Hiroaki; Okubo, Masaaki; Igashira, Masayuki [Tokyo Inst. of Tech. (Japan); Horiki, Oichiro; Suzaki, Takenori

1997-10-01

The Tank-Type Critical Assembly (TCA) of Japan Atomic Energy Research Institute is research equipment for light water reactor physics. In the present report, the lectures given to the graduate students of Tokyo Institute of Technology who participated in the educational experiment course held on 26-30 August at TCA are rearranged to provide useful information for those who will implement educational basic experiments with TCA in the future. This report describes the principles, procedures, and data analyses for (1) Critical approach and Exponential experiment, (2) Measurement of neutron flux distribution, (3) Measurement of power distribution, (4) Measurement of fuel rod worth distribution, and (5) Measurement of safety plate worth by the rod drop method. (author)

Educational reactor-physics experiments with the critical assembly TCA

International Nuclear Information System (INIS)

Tsutsui, Hiroaki; Okubo, Masaaki; Igashira, Masayuki; Horiki, Oichiro; Suzaki, Takenori.

1997-10-01

The Tank-Type Critical Assembly (TCA) of Japan Atomic Energy Research Institute is research equipment for light water reactor physics. In the present report, the lectures given to the graduate students of Tokyo Institute of Technology who participated in the educational experiment course held on 26-30 August at TCA are rearranged to provide useful information for those who will implement educational basic experiments with TCA in the future. This report describes the principles, procedures, and data analyses for 1) Critical approach and Exponential experiment, 2) Measurement of neutron flux distribution, 3) Measurement of power distribution, 4) Measurement of fuel rod worth distribution, and 5) Measurement of safety plate worth by the rod drop method. (author)

Monte Carlo Depletion with Critical Spectrum for Assembly Group Constant Generation

International Nuclear Information System (INIS)

Park, Ho Jin; Joo, Han Gyu; Shim, Hyung Jin; Kim, Chang Hyo

2010-01-01

The conventional two-step procedure has been used in practical nuclear reactor analysis. In this procedure, a deterministic assembly transport code such as HELIOS and CASMO is normally to generate multigroup flux distribution to be used in few-group cross section generation. Recently there are accuracy issues related with the resonance treatment or the double heterogeneity (DH) treatment for VHTR fuel blocks. In order to mitigate the accuracy issues, Monte Carlo (MC) methods can be used as an alternative way to generate few-group cross sections because the accuracy of the MC calculations benefits from its ability to use continuous energy nuclear data and detailed geometric information. In an earlier work, the conventional methods of obtaining multigroup cross sections and the critical spectrum are implemented into the McCARD Monte Carlo code. However, it was not complete in that the critical spectrum is not reflected in the depletion calculation. The purpose of this study is to develop a method to apply the critical spectrum to MC depletion calculations to correct for the leakage effect in the depletion calculation and then to examine the MC based group constants within the two-step procedure by comparing the two-step solution with the direct whole core MC depletion result

Measurement of critical mass for an assembly of bare uranium shells

International Nuclear Information System (INIS)

Myers, W.L.; Goulding, C.A.; Hollas, C.L.

1997-01-01

As part of the research into nuclear measurement techniques, a series of measurements was performed that have applications to criticality safety and nuclear material handling. The critical mass of a set of bare, enriched-uranium metal hemispherical shells, known as the Rocky Flats shells, was measured for an assembly having an inside radius of 2.347 cm. The critical mass value was extrapolated from a series of subcritical measurements using three different kinds of sources (AmBe, AmF, and 252 Cf) placed at the center of the shells. Two kinds of neutron detection configurations (a 1% efficiency and a 25% efficiency configuration) were used to make the measurements

Critical assembly of uranium enriched to 10% in uranium-235

International Nuclear Information System (INIS)

Hansen, G.E.; Paxton, H.E.

1979-01-01

Big Ten is described in the detail appropriate for a benchmark critical assembly. Characteristics provided are spectral indexes and a detailed neutron flux spectrum, Rossi-α on a reactivity scale established by positive periods, and reactivity coefficients of a variety of isotopes, including the fissionable materials. The observed characteristics are compared with values calculated with ENDF/B-IV cross sections

Benchmarking of HEU mental annuli critical assemblies with internally reflected graphite cylinder

Directory of Open Access Journals (Sweden)

Xiaobo Liu

2017-01-01

Full Text Available Three experimental configurations of critical assemblies, performed in 1963 at the Oak Ridge Critical Experiment Facility, which are assembled using three different diameter HEU annuli (15-9 inches, 15-7 inches and 13-7 inches metal annuli with internally reflected graphite cylinder are evaluated and benchmarked. The experimental uncertainties which are 0.00057, 0.00058 and 0.00057 respectively, and biases to the benchmark models which are − 0.00286, − 0.00242 and − 0.00168 respectively, were determined, and the experimental benchmark keff results were obtained for both detailed and simplified models. The calculation results for both detailed and simplified models using MCNP6-1.0 and ENDF/B-VII.1 agree well to the benchmark experimental results within difference less than 0.2%. The benchmarking results were accepted for the inclusion of ICSBEP Handbook.

Analysis of measurements for a uranium-free core experiment at the BFS-2 critical assembly

Energy Technology Data Exchange (ETDEWEB)

Hunter, Stuart [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

1999-04-01

This document describes a series of calculations that were carried out to model various measurements from the BFS-58-1-I1 experiment. BFS-58-1-I1 was a mock-up of a uranium-free, Pu burning core at BFS-2, a Russian critical assembly operated by IPPE. The experiment measured values of Keff, Na void reactivity worth, material sample reactivity worths and reaction rate ratios. The experiments were modelled using a number of different methods. Basic nuclear data was taken from JENDL-3.2, in either 70 or 18 groups. Cross-section data for the various material regions of the assembly were calculated by either SLAROM or CASUP; the heterogeneous structure of the core regions was modelled in these calculations, with 3 different options considered for representing the (essentially 2D) geometry of the assembly components in a 1D cell model. Whole reactor calculations of flux and Keff were done using both a diffusion model (CITATION) and a transport model (TWOTRAN2), both using an RZ geometry. Reactivity worths were calculated both directly from differences in Keff values and by using the exact perturbation calculations of PERKY and SN-PERT (for CITATION and TWOTRAN2, respectively). Initial calculations included a number of inaccuracies in the assembly representation, a result of communication difficulties between JNC and IPPE; these errors were removed for the final calculations that are presented. Calculations for the experiments have also been carried out in Russia (IPPE) and France (CEA) as part of an international comparison exercise, some of those results are also presented here. The calculated value of Keff was 1.1%{delta}k/k higher than the measured value, Na void worth C/E values were {approx}1.06; these results were considered to be reasonable. (Discrepancies in certain Na void values were probably due to experimental causes , though the effect should be investigated in any future experiments.) Several sample worth values were small compared with calculational

Fellowship at orita: A critical analysis of the leadership crisis in the Assemblies of God, Nigeria

Directory of Open Access Journals (Sweden)

Williams O. Mbamalu

2016-07-01

Full Text Available This article is a critical analysis of the present crisis in the Assemblies of God, Nigeria (AGN. A background history of the church is given to show how growth had taken place and how decline had set in. Doing this involves analysing the factors responsible for the present crisis that has brought the church to its knees. The article finds that the AGN’s membership and leadership are dominated by the Igbo ethnic group whose worldviews are known to be highly competitive, individualistic and ‘pantomimic’. The AGN’s constitution and bye-laws do not include a clause that prevents pastors from the same ethnic group from holding the two top-most positions of the General Superintendent and the Assistant General Superintendent at the same time. Therefore the article submits that the AGN should amend its constitution to deal with these pertinent issues. The significance of the article is that it calls the attention of other Pentecostal denominations in Nigeria and the rest of Africa to the crisis-ridden AGN, whose eschatological and Pentecostal persuasion is at orita [the crossroads] and urges them to learn from it.

Consideration of criticality in a nuclear waste repository

International Nuclear Information System (INIS)

Rechard, R.P.; Sanchez, L.C.; Stockman, C.T.; Ramsey, J.L. Jr.; Martell, M.

1995-01-01

The preliminary criticality analysis that was done suggests that the possibility of achieving critical conditions cannot be easily ruled out without looking at the geochemical process of assembly or the dynamics of the operation of a critical assembly. The evaluation of a critical assembly requires an integrated, consistent approach that includes evaluating the following: (1) the alteration rates of the layers of the container and spent fuel, (2) the transport of fissile material or neutron absorbers, and (3) the assembly mechanisms that can achieve critical conditions. The above is a non-trivial analysis and preliminary work suggests that with the loading assumed, enough fissile mass will leach from the HEU multi-purpose canisters to support a criticality. In addition, the consequences of an unpressurized Oklo type criticality would be insignificant to the performance of an unsaturated, tuff repository

Boiling transition phenomenon in BWR fuel assemblies effect of fuel spacer shape on critical power

International Nuclear Information System (INIS)

Yamamoto, Yasushi; Morooka, Shin-ichi; Mitsutake, Toru; Yokobori, Seiichi; Kimura, Jiro.

1996-01-01

A thorough understanding of the thermal-hydraulic phenomena near fuel spacer is necessary for the accurate prediction of the critical power of BWR fuel assemblies, and is thus essential for effective developments of a new BWR fuel assembly. The main purpose of this study is to develop an accurate method for predicting the effect of spacer shapes on critical power. Tests have been conducted under actual BWR operating conditions, using an annulus flow channel consisting of a heated rod and circular-tube channel, and BWR simulated 4x4 rod bundles with heater rods unheated just upsteam of spacer. The effect of spacer shapes on critical power was predicted analytically based on the droplet deposition rate estimation. The droplet deposition rate for different spacer shapes was calculated using a single-phase flow model. The prediction results were compared with the test results for the annulus flow channel using ring-type spacers. Analytical results of critical power agreed with measured critical power from point of the effects of changes in the rod-spacer clearance and the spacer thickness on critical power. (author)

Criticality safety assessment of FBTR fuel sub-assemblies using WIMS cross section set

International Nuclear Information System (INIS)

Gupta, H.C.; Chakraborty, B.

2002-01-01

Full text: FBTR's irradiated fuel sub-assemblies (FSAs) are sent to RML at Indira Gandhi Centre for Atomic Research for post irradiation examination. The FSAs are cut open and the fuel pins are separated for examination in the hot cells. It was required to evaluate the criticality safety in handling the FSAs in the hot cells. Criticality safety studies for handling two as well as three irradiated FSAs in the hot cells under dry conditions were carried out by the Safety Group at IGCAR, Kalpakkam. Monte Carlo code KENO (Version Va) which uses 16-group Hansen-Roach cross-section set was used for the calculations. Subsequently, during the safety review of the proposition by the Safety Review Committee (SARCOP) of AERB, it was stipulated to carry out the criticality safety studies under flooded condition also. We carried out the criticality safety studies for these fuel sub assemblies in different configurations under dry (buried in concrete) as well as wet condition (flooded with light water) using Monte Carlo codes MONALI (developed at BARC) and KENO4 using WlMS-69 group cross section set. Results of our analyses under various conditions are presented in this paper

A simplified treatment of radial enrichment distributions of LWR fuel assemblies in criticality calculations

International Nuclear Information System (INIS)

Hennebach, M.; Schnorrenberg, N.

2008-01-01

Criticality safety assessments are usually performed for fuel assembly models that are as generic as possible to encompass small modifications in geometry that have no impact on criticality. Dealing with different radial enrichment distributions for a fuel assembly type, which is especially important for BWR fuel, poses more of a challenge, since this characteristic is rather obviously influencing the neutronic behaviour of the system. Nevertheless, the large variability of enrichment distributions makes it very desirable and even necessary to treat them in a generalized way, both to keep the criticality safety assessment from becoming too unwieldy and to avoid having to extend it every time a new variation comes up. To be viable, such a generic treatment has to be demonstrably covering, i.e. lead to a higher effective neutron multiplication factor k eff than any of the radial enrichment distributions it represents. Averaging the enrichment evenly over the fuel rods of the assembly is a general and simple approach, and under reactor conditions, it is also a covering assumption: the graded distribution is introduced to achieve a linear power distribution, therefore reducing the enrichment of the better moderated rods at the edge of the assembly. With an even distribution of the average enrichment over all rods, these wellmoderated rods will cause increased fission rates at the assembly edges and a rise in k eff . Since the moderator conditions in a spent nuclear fuel cask differ strongly from a reactor even when considering optimal moderation, the proof that a uniform enrichment distribution is a covering assumption compared with detailed enrichment distributions has to be cask-specific. In this report, a method for making that proof is presented along with results for fuel assemblies from BWR reactors. All results are from three-dimensional Monte Carlo calculations with the SCALE 5.1 code package [1], using a 44-group neutron crosssection library based on ENDF

Mechanics of a crushable pebble assembly using discrete element method

International Nuclear Information System (INIS)

Annabattula, R.K.; Gan, Y.; Zhao, S.; Kamlah, M.

2012-01-01

The influence of crushing of individual pebbles on the overall strength of a pebble assembly is investigated using discrete element method. An assembly comprising of 5000 spherical pebbles is assigned with random critical failure energies with a Weibull distribution in accordance with the experimental observation. Then, the pebble assembly is subjected to uni-axial compression (ε 33 =1.5%) with periodic boundary conditions. The crushable pebble assembly shows a significant difference in stress–strain response in comparison to a non-crushable pebble assembly. The analysis shows that a ideal plasticity like behaviour (constant stress with increase in strain) is the characteristic of a crushable pebble assembly with sudden damage. The damage accumulation law plays a critical role in determining the critical stress while the critical number of completely failed pebbles at the onset of critical stress is independent of such a damage law. Furthermore, a loosely packed pebble assembly shows a higher crush resistance while the critical stress is insensitive to the packing factor (η) of the assembly.

Thermal adaptation of mesophilic and thermophilic FtsZ assembly by modulation of the critical concentration.

Directory of Open Access Journals (Sweden)

Luis Concha-Marambio

Full Text Available Cytokinesis is the last stage in the cell cycle. In prokaryotes, the protein FtsZ guides cell constriction by assembling into a contractile ring-shaped structure termed the Z-ring. Constriction of the Z-ring is driven by the GTPase activity of FtsZ that overcomes the energetic barrier between two protein conformations having different propensities to assemble into polymers. FtsZ is found in psychrophilic, mesophilic and thermophilic organisms thereby functioning at temperatures ranging from subzero to >100°C. To gain insight into the functional adaptations enabling assembly of FtsZ in distinct environmental conditions, we analyzed the energetics of FtsZ function from mesophilic Escherichia coli in comparison with FtsZ from thermophilic Methanocaldococcus jannaschii. Presumably, the assembly may be similarly modulated by temperature for both FtsZ orthologs. The temperature dependence of the first-order rates of nucleotide hydrolysis and of polymer disassembly, indicated an entropy-driven destabilization of the FtsZ-GTP intermediate. This destabilization was true for both mesophilic and thermophilic FtsZ, reflecting a conserved mechanism of disassembly. From the temperature dependence of the critical concentrations for polymerization, we detected a change of opposite sign in the heat capacity, that was partially explained by the specific changes in the solvent-accessible surface area between the free and polymerized states of FtsZ. At the physiological temperature, the assembly of both FtsZ orthologs was found to be driven by a small positive entropy. In contrast, the assembly occurred with a negative enthalpy for mesophilic FtsZ and with a positive enthalpy for thermophilic FtsZ. Notably, the assembly of both FtsZ orthologs is characterized by a critical concentration of similar value (1-2 μM at the environmental temperatures of their host organisms. These findings suggest a simple but robust mechanism of adaptation of FtsZ, previously shown

CSER-00-007 Addendum 1 Criticality Safety Evaluation of Shippingport PWR Core 2 Blanket Fuel Assemblies at Lower Exposures

International Nuclear Information System (INIS)

WITTEKIND, W.D.

2001-01-01

This analysis meets the requirements of HNF-7098, Criticality Safety Program, (FH 2001a). HNF-7098 states that before starting a new operation with fissile material or before an existing operation is changed, it shall be determined that the entire process will be subcritical under both normal and credible abnormal conditions. To demonstrate the Incredibility Principle is satisfied, this Criticality Safety Evaluation Report (CSER) shows that the form or distribution is such that criticality is impossible. This evaluation demonstrated, that on the basis of effective 235 U enrichment, criticality is not possible. The minimum blanket assembly exposure is 4,375 MW t d/MTU for fissile material that is shown to fulfill the Incredibility Principle safety criterion on the basis of enrichment

Shippingport LWBR (Th/U Oxide) Fuel Characteristics for Disposal Criticality Analysis

International Nuclear Information System (INIS)

Taylor, L. L.; Loo, H. H.

1999-01-01

Department of Energy (DOE)-owned spent nuclear fuels encompass many fuel types. In an effort to facilitate criticality analysis for these various fuel types, they were categorized into eight characteristic fuel groups with emphasis on fuel matrix composition. Out of each fuel group, a representative fuel type was chosen for analysis as a bounding case within that fuel group. Generally, burnup data, fissile enrichments, and total fuel and fissile mass govern the selection of the representative or candidate fuel within that group. The Shippingport Light Water Breeder Reactor (LWBR) fuels incorporate more of the conventional materials (zirconium cladding/heavy metal oxides) and fabrication details (rods and spacers) that make them comparable to a typical commercial fuel assembly. The LWBR seed/blanket configuration tested a light-water breeder concept with Th-232/U-233 binary fuel matrix. Reactor design used several assembly configurations at different locations within the same core . The seed assemblies contain the greatest fissile mass per (displaced) unit volume, but the blanket assemblies actually contain more fissile mass in a larger volume; the atom-densities are comparable

Vibration characteristics analysis for HANARO fuel assembly

International Nuclear Information System (INIS)

Ryu, Jeong Soo; Yoon, Doo Byung

2001-06-01

For investigating the vibration characteristics of HANARO fuel assembly, the finite element models of the in-air fuel assemblies and flow tubes were developed. By calculating the hydrodynamic mass and distributing it on the in-air models, the in-water models of the flow tubes and the fuel assemblies were developed. Then, modal analysis of the developed models was carried out. The analysis results show that the fundamental vibration modes of the in-air 18-element and 36-element fuel assemblies are lateral bending modes and its corresponding natural frequencies are 26.4Hz and 27.7Hz, respectively. The fundamental natural frequency of the in-water 18-element and 36-element fuel assemblies were obtained as 16.1Hz and 16.5Hz. For the verification of the developed finite element models, modal analysis results were compared with those obtained from the modal test. These results demonstrate that the natural frequencies of lower order modes obtained from finite element analysis agree well with those of the modal test and the estimation of the hydrodynamic mass is appropriate. It is expected that the analysis results will be applied as a basic data for the operation and management of the HANARO. In addition, when it is necessary to improve the design of the fuel assembly, the developed finite element models will be utilized as a base model for the vibration characteristic analysis of the modified fuel assembly

Data base of reactor physics experimental results in Kyoto University critical assembly experimental facilities

International Nuclear Information System (INIS)

Ichihara, Chihiro; Fujine, Shigenori; Hayashi, Masatoshi

1986-01-01

The Kyoto University critical assembly experimental facilities belong to the Kyoto University Research Reactor Institute, and are the versatile critical assembly constructed for experimentally studying reactor physics and reactor engineering. The facilities are those for common utilization by universities in whole Japan. During more than ten years since the initial criticality in 1974, various experiments on reactor physics and reactor engineering have been carried out using many experimental facilities such as two solidmoderated cores, a light water-moderated core and a neutron generator. The kinds of the experiment carried out were diverse, and to find out the required data from them is very troublesome, accordingly it has become necessary to make a data base which can be processed by a computer with the data accumulated during the past more than ten years. The outline of the data base, the data base CAEX using personal computers, the data base supported by a large computer and so on are reported. (Kako, I.)

Criticality Analysis of SFP Region I under Dry Air Condition

International Nuclear Information System (INIS)

Kim, Ki Yong; Kim, Min Chul

2016-01-01

This paper is to provide a result of the criticality evaluation under the condition that new fuel assemblies for initial fuel loading are storing in Region 1 of SFP in the dry air. The objective of this analysis is to ensure the effective neutron multiplication factor(k_e_f_f) of SFP is less than 0.95 under that condition. This analysis ensured the effective neutron multiplication factor(k_e_f_f) of Region 1 of SFP is less than 0.95 under the condition in the air. The keff in Region I of SFP under the condition of the dry air is 0.5865. The increased k_c_a_l_c of the Region 1 after the mislocated fuel assembly accident is 0.0444 at the pool flooded with un-borated water

A comparison of reaction rate calculations using Endf/B-VII with critical assembly measurements

International Nuclear Information System (INIS)

Wilkerson, C.; Mac Innes, M.; Barr, D.; Trellue, H.; MacFarlane, R.; Chadwick, M.

2008-01-01

We present critical assembly reaction rate data, and modeling of the same using the recently released Endf/B-VII library. While some of the experimental measurements were performed as long as 50 years ago, the results have not been widely used/available outside of Los Alamos. Over the years, a variety of target foils were fabricated and placed in differing neutron spectrum/fluence environments within critical assemblies. Neutron-induced reactions such as (n,γ), (n,2n), and (n,f) on these targets were measured, typically referenced to 235 U(n,f) or 239 Pu(n,f). Because the cross section for the latter reactions are now well known, these experiments provide a rich data set for testing evaluated cross sections. Due to the large variety of critical assemblies that were historically available at Los Alamos, it was possible to make measurements in spectral environments ranging from hard (Pu Jezebel, center of Pu Flattop) through intermediate (Big Ten) to degraded (reflector region of Flattop). This broad range of configurations allows us to test both the cross section magnitudes and their energy dependencies. We will present data, along with reaction rate predictions using primarily MCNP5 in conjunction with Endf/B-VII, for a number of target nuclei, including iridium, isotopes of uranium (e.g., 233, 235, 237, 238), neptunium (237), plutonium (239), and americium (241). (authors)

Benchmarks of subcriticality in accelerator-driven system at Kyoto University Critical Assembly

Directory of Open Access Journals (Sweden)

Cheol Ho Pyeon

2017-09-01

Full Text Available Basic research on the accelerator-driven system is conducted by combining 235U-fueled and 232Th-loaded cores in the Kyoto University Critical Assembly with the pulsed neutron generator (14 MeV neutrons and the proton beam accelerator (100 MeV protons with a heavy metal target. The results of experimental subcriticality are presented with a wide range of subcriticality level between near critical and 10,000 pcm, as obtained by the pulsed neutron source method, the Feynman-α method, and the neutron source multiplication method.

Coupled neutronics/thermal-hydraulics analysis of a high-performance light-water reactor fuel assembly

International Nuclear Information System (INIS)

Waata, C.L.

2006-07-01

The use of water at supercritical pressure as coolant and moderator introduces a challenge in the design of a High-Performance Light-Water Reactor (HPLWR) fuel assembly. At supercritical pressure condition (P=25 MPa), the thermal-hydraulics behaviour of water differs strongly from that at sub-critical pressure due to a rapid variation of the thermal-physical properties across the pseudo-critical line. Due of the strong link between the water (moderation) and the neutron spectrum and subsequently the power distribution, a coupling of neutronics and thermal-hydraulics has become a necessity for reactor concepts operating at supercritical pressure condition. The effect of neutron moderation on the local parameters of thermal-hydraulics and vice-verse in a fuel assembly has to be considered for an accurate design analysis. In this study, the Monte Carlo N-Particle code (MCNP) and the sub-channel code STAFAS (Sub-channel Thermal-hydraulics Analysis of a Fuel Assembly under Supercritical conditions) have been coupled for the design analysis of a fuel assembly with supercritical water as coolant and moderator. Both codes are well known for complex geometry modelling. The MCNP code is used for neutronics analyses and for the prediction of power profiles of individual fuel rods. The sub-channel code STAFAS for the thermal-hydraulics analyses takes into account the coolant properties beyond the critical point as well as separate moderator channels. The coupling procedure is realized automatically. MCNP calculates the power distribution in each fuel rod, which is then transferred into STAFAS to obtain the corresponding thermal-hydraulic conditions in each sub-channel. The new thermal-hydraulic conditions are used to generate a new input deck for the next MCNP calculation. This procedure is repeated until a converged state is achieved. The coupled code system was tested on a proposed fuel assembly design of a HPLWR. An under-relaxation was introduced to achieve convergence

The Prompt Fission Neutron Spectrum: From Experiment to the Evaluated Data and its Impact on Critical Assemblies

Energy Technology Data Exchange (ETDEWEB)

Rising, Michael Evan [Los Alamos National Laboratory

2015-06-10

After a brief introduction concerning nuclear data, prompt fission neutron spectrum (PFNS) evaluations and the limited PFNS covariance data in the ENDF/B-VII library, and the important fact that cross section uncertainties ~ PFNS uncertainties, the author presents background information on the PFNS (experimental data, theoretical models, data evaluation, uncertainty quantification) and discusses the impact on certain well-known critical assemblies with regard to integral quantities, sensitivity analysis, and uncertainty propagation. He sketches recent and ongoing research and concludes with some final thoughts.

Basket criticality design of a dual purpose cask for VVER 1000 spent fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Rezaeian, Mahdi [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Kamali, Jamshid [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of)

2016-12-15

Dual purpose cask technology is one of the most prominent options for interim storage of spent fuels following their removal from reactors. Criticality safety of the spent fuel assemblies are ensured by design of the basket within these casks. In this study, a set of criticality design calculations of a dual purpose cask for 12 VVER 1000 spent fuel assemblies of Bushehr nuclear power plant were carried out. The basket material of borated stainless steel with 0.5 to 2.5 wt% of boron and Boral (Al-B{sub 4}C) with 1.5 to 40 wt% of boron carbide, were investigated and the minimum required receptacle pitch of the basket was determined. Using the calculated receptacle pitch of the basket, the minimum required diameter of the cavity could be established.

A new facility for the determination of critical heat flux in nuclear fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Fortman, R A; Hadaller, G I; Hamilton, R C; Hayes, R C; Shin, K S; Stern, F [Stern Laboratories Inc., Hamilton, ON (Canada)

1993-11-01

A facility for the determination of critical heat flux in simulated reactor fuel assemblies has been constructed at Stern Laboratories for CANDU Owners` Group. This paper describes the facility and method of testing. 9 figs.

Structural analysis of ITER sub-assembly tools

International Nuclear Information System (INIS)

Nam, K.O.; Park, H.K.; Kim, D.J.; Ahn, H.J.; Lee, J.H.; Kim, K.K.; Im, K.; Shaw, R.

2011-01-01

The ITER Tokamak assembly tools are purpose-built assembly tools to complete the ITER Tokamak machine which includes the cryostat and the components contained therein. The sector sub-assembly tools descried in this paper are main assembly tools to assemble vacuum vessel, thermal shield and toroidal filed coils into a complete 40 o sector. The 40 o sector sub-assembly tools are composed of sector sub-assembly tool, including radial beam, vacuum vessel supports and mid-plane brace tools. These tools shall have sufficient strength to transport and handle heavy weight of the ITER Tokamak machine reached several hundred tons. Therefore these tools should be designed and analyzed to confirm both the strength and structural stability even in the case of conservative assumptions. To verify structural stabilities of the sector sub-assembly tools in terms of strength and deflection, ANSYS code was used for linear static analysis. The results of the analysis show that these tools are designed with sufficient strength and stiffness. The conceptual designs of these tools are briefly described in this paper also.

Analysis of Human Communication during Assembly Tasks.

Science.gov (United States)

1986-06-01

AD-A7l 43 ANALYSIS OF HUMAN COMMUNICATION DURING ASSEMBLY TASKS in1(U) CRNEGIE-MELLO UNIY PITTSBURGH PA ROBOTICS INST UNCLSSIIEDK S BARBER ET AL...ao I Dur~~~~IngAbcbyTs; 7c .S:in i lSAo .0. Analysis of Human Communication During Assembly Tasks K. Suzanne Barber and Gerald J. Agin CMU-RI-TR-86-1...TYPE or REPORT & PE-Rioo CevCZaz Analysis of Human Communication During Assembly Inlterim Tasks I . PERFORMING 00RG. REPORT NUMBER 1. £UT~oOR~e) IL

Water/sand flooded and immersed critical experiment and analysis performed in support of the TOPAZ-II Safety Program

International Nuclear Information System (INIS)

Glushkov, E.S.; Ponomarev-Stepnoi, N.N.; Bubelev, V.G.; Garin, V.P.; Gomin, E.A.; Kompanietz, G.V.; Krutoy, A.M.; Lobynstev, V.A.; Maiorov, L.V.; Polyakov, D.N.

1994-01-01

Presented is a brief description of the Narciss-M2 critical assemblies, which simulate accidental water/wet-sand immersion of the TOPAZ-II reactor as well as water-flooding of core cavities. Experimental results obtained from these critical assemblies, including experiments with several fuel elements removed from the core, are shown. These configurations with several extracted fuel elements simulate a proposed fuel-out anticriticality-device modification to the TOPAZ-II reactor. Preliminary computational analysis of these experiments using the Monte Carlo neutron-transport method is outlined. Nuclear criticality safety of the TOPAZ-II reactor with an incorporated anticriticality unit is demonstrated

Spool assembly support analysis

International Nuclear Information System (INIS)

Norman, B.F.

1994-01-01

This document provides the wind/seismic analysis and evaluation for the pump pit spool assemblies. Hand calculations were used for the analysis. UBC, AISC, and load factors were used in this evaluation. The results show that the actual loads are under the allowable loads and all requirements are met

Safe operation of critical assemblies and research reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1960-09-15

Some countries have accumulated considerable experience in the operation of these reactors and have in the process developed safe practices. On the other hand, other countries which have recently acquired, or will soon acquire, such reactors do not have sufficient background of experience with them to have developed full knowledge regarding their safe operation. In this situation, the International Atomic Energy Agency has considered that it would be useful to make available to all its Member States a set of recommendations on the safe operation of these reactors, based on the accumulated experience and best practices. The Director General accordingly nominated a Pane Ion Safe Operation of Critical Assemblies and Research Reactors to assist the Agency's Secretariat in drafting such recommendations

Analysis of PWR assembly bow

International Nuclear Information System (INIS)

Fetterman, Robert J.; Franceschini, Fausto

2008-01-01

Excessive out of core assembly bow has been observed during refueling outages of certain PWRs. Assembly bow can take on a rather complex S-shape, and in other cases C-shape bow is prevalent. Concerns have been raised regarding the impact of the observed assembly bow on the in-core power distribution and the safety analyses supporting the plant operations. In response to these concerns, Westinghouse has developed a comprehensive analysis process for determining the effects of assembly bow on core power distributions and plant operating margins. This methodology has been applied to a particular reactor as part of an overall safety reanalysis completed in support of plant modifications. This paper provides a brief description of the methods used and a summary of the pertinent results. (authors)

Analysis of PWR assembly bow

Energy Technology Data Exchange (ETDEWEB)

Fetterman, Robert J.; Franceschini, Fausto [Westinghouse Electric Company LLC, Pittsburgh, PA (United States)

2008-07-01

Excessive out of core assembly bow has been observed during refueling outages of certain PWRs. Assembly bow can take on a rather complex S-shape, and in other cases C-shape bow is prevalent. Concerns have been raised regarding the impact of the observed assembly bow on the in-core power distribution and the safety analyses supporting the plant operations. In response to these concerns, Westinghouse has developed a comprehensive analysis process for determining the effects of assembly bow on core power distributions and plant operating margins. This methodology has been applied to a particular reactor as part of an overall safety reanalysis completed in support of plant modifications. This paper provides a brief description of the methods used and a summary of the pertinent results. (authors)

Computer Tomography Analysis of Fastrac Composite Thrust Chamber Assemblies

Science.gov (United States)

Beshears, Ronald D.

2000-01-01

Computed tomography (CT) inspection has been integrated into the production process for NASA's Fastrac composite thrust chamber assemblies (TCAs). CT has been proven to be uniquely qualified to detect the known critical flaw for these nozzles, liner cracks that are adjacent to debonds between the liner and overwrap. CT is also being used as a process monitoring tool through analysis of low density indications in the nozzle overwraps. 3d reconstruction of CT images to produce models of flawed areas is being used to give program engineers better insight into the location and nature of nozzle flaws.

A criticality analysis of the GBC-32 dry storage cask with Hanbit nuclear power plant unit 3 fuel assemblies from the viewpoint of burnup credit

Energy Technology Data Exchange (ETDEWEB)

Yun, Hyung Ju; Kim, Do Yeon; Park, Kwang Heon; Hong, Ser Gi [Dept. of Nuclear Engineering, Kyung Hee University, Seoul (Korea, Republic of)

2016-06-15

Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that keff values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.

Water/sand flooded and immersed critical experiment and analysis performed in support of the TOPAZ-II safety program

International Nuclear Information System (INIS)

Glushkov, E.S.; Ponomarev-Stepnoi, N.N.; Bubelev, V.G.; Garin, V.P.; Gomin, E.A.; Kompanietz, G.V.; Krutov, A.M.; Lobynstev, V.A.; Maiorov, L.V.; Polyakov, D.N.; Chunyaev, E.I.; Marshall, A.C.; Sapir, J.L.; Pelowitz, D.B.

1995-01-01

Presented is a brief description of the Narciss-M2 critical assemblies, which simulate accidental water/wet-sand immersion of the TOPAZ-II reactor as well as water-flooding of core cavities. Experimental results obtained from these critical assemblies, including experiments with several fuel elements removed from the core, are shown. These configurations with several extracted fuel elements simulate a proposed fuel-out anticriticality-device modification to the TOPAZ-II reactor. Preliminary computational analysis of these experiments using the Monte Carlo neutron-transport method is outlined. Nuclear criticality safety of the TOPAZ-II reactor with an incorporated anticriticality unit is demonstrated. copyright 1995 American Institute of Physics

ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

Energy Technology Data Exchange (ETDEWEB)

Kahler, A. [Los Alamos National Laboratory (LANL); Macfarlane, R E [Los Alamos National Laboratory (LANL); Mosteller, R D [Los Alamos National Laboratory (LANL); Kiedrowski, B C [Los Alamos National Laboratory (LANL); Frankle, S C [Los Alamos National Laboratory (LANL); Chadwick, M. B. [Los Alamos National Laboratory (LANL); Mcknight, R D [Argonne National Laboratory (ANL); Lell, R M [Argonne National Laboratory (ANL); Palmiotti, G [Idaho National Laboratory (INL); Hiruta, h [Idaho National Laboratory (INL); Herman, Micheal W [Brookhaven National Laboratory (BNL); Arcilla, r [Brookhaven National Laboratory (BNL); Mughabghab, S F [Brookhaven National Laboratory (BNL); Sublet, J C [Culham Science Center, Abington, UK; Trkov, A. [Jozef Stefan Institute, Slovenia; Trumbull, T H [Knolls Atomic Power Laboratory; Dunn, Michael E [ORNL

2011-01-01

The ENDF/B-VII.1 library is the latest revision to the United States' Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 423 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [1]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Project's International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unrnoderated and uranium reflected (235)U and (239)Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also confirmed for selected actinide reaction rates such as (236)U; (238,242)Pu and (241,243)Am capture in fast systems. Other deficiencies, such as the overprediction of Pu solution system critical

Research on the reactor physics using the Kyoto University Critical Assembly (KUCA)

International Nuclear Information System (INIS)

1986-10-01

The Kyoto University Critical Assembly [KUCA] is a multi-core type critical assembly established in 1974, as a facility for the joint use study by researchers of all universities in Japan. Thereafter, many reactor physics experiments have been carried out using three cores (A-, B-, and C-cores) in the KUCA. In the A- and B-cores, solid moderator such as polyethylene or graphite is used, whereas light-water is utilized as moderator in the C-core. The A-core has been employed mainly in connection with the Cockcroft-Walton type accelerator installed in the KUCA, to measure (1) the subcriticality by the pulsed neutron technique for the critical safety research and (2) the neutron spectrum by the time-of-flight technique. Recently, a basic study on the tight lattice core has also launched using the A-core. The B-core has been employed for the research on the thorium fuel cycle ever since. The C-core has been employed (1) for the basic studies on the nuclear characteristics of light-water moderated high-flux research reactors, including coupled-cores, and (2) for a research related to reducing enrichment of uranium fuel used in research reactors. The C-core is being utilized in the reactor laboratory course experiment for students of ten universities in Japan. The data base of the KUCA critical experiments is generated so far on the basis of approximately 350 experimental reports accumulated in the KUCA. Besides, the assessed KUCA code system has been established through analyses on the various KUCA experiments. In addition to the KUCA itself, both of them are provided for the joint use study by researchers of all universities in Japan. (author)

Risk management for operations of the LANL Critical Experiments Facility

International Nuclear Information System (INIS)

Paternoster, R.; Butterfield, K.

1998-01-01

The Los Alamos Critical Experiments Facility (LACEF) currently operates two burst reactors (Godiva-IV and Skua), one solution assembly [the Solution High-Energy Burst Assembly (SHEBA)], two fast-spectrum benchmark assemblies (Flattop and Big Ten), and five general-purpose remote assembly machines that may be configured with nuclear materials and assembled by remote control. Special nuclear materials storage vaults support these and other operations at the site. With this diverse set of operations, several approaches are possible in the analysis and management of risk. The most conservative approach would be to write a safety analysis report (SAR) for each assembly and experiment. A more cost-effective approach is to analyze the probability and consequences of several classes of operations representative of operations on each critical assembly machine and envelope the bounding case accidents. Although the neutron physics of these machines varies widely, the operations performed at LACEF fall into four operational modes: steady-state mode, approach-to-critical mode, prompt burst mode, and nuclear material operations, which can include critical assembly fuel loading. The operational sequences of each mode are very nearly identical, whether operated on one assembly machine or another. The use of an envelope approach to accident analysis is facilitated by the use of classes of operations and the use of bounding case consequence analysis. A simple fault tree analysis of operational modes helps resolve which operations are sensitive to human error and which are initiated by hardware of software failures. Where possible, these errors and failures are blocked by TSR LCOs. Future work will determine the probability of accidents with various initiators

Risk management for operations of the Los Alamos critical experiments facility

International Nuclear Information System (INIS)

Paternoster, R.; Butterfield, K.

1998-01-01

The Los Alamos Critical Experiments Facility (LACEF) currently operates two burst reactors (Godiva-IV and Skua), one solution assembly (SHEBA 2--Solution high-Energy Burst Assembly), two fast-spectrum benchmark assemblies (Flattop and Big Ten), and five general-purpose remote assembly machines which may be configured with nuclear materials and assembled by remote control. SNM storage vaults support these and other operations at the site. With this diverse set of operations, several approaches are possible in the analysis and management of risk. The most conservative approach would be to write a safety analysis report (SAR) for each assembly and experiment. A more cost-effective approach is to analyze the probability and consequences of several classes of operations representative of operations on each critical assembly machine and envelope the bounding case accidents. Although the neutron physics of these machines varies widely, the operations performed at LACEF fall into four operational modes: steady-state mode, approach-to-critical mode, prompt burst mode, and nuclear material operations which can include critical assembly fuel loading. The operational sequences of each mode are very nearly the same, whether operated on one assembly machine or another. The use of an envelope approach to accident analysis is facilitated by the use of classes of operations and the use of bounding case consequence analysis. A simple fault tree analysis of operational modes helps resolve which operations are sensitive to human error and which are initiated by hardware of software failures. Where possible, these errors and failures are blocked by TSR LCOs

Criticality alarm device

International Nuclear Information System (INIS)

Kasai, Kenji.

1994-01-01

The device of the present invention is utilized, for example, to a reprocessing facility for storing and processing nuclear fuels and measures and controls the nuclear fuel assembly system so as not to exceed criticality. That is, a conventional criticality alarm device applies a predetermined processing to neutron fluxes generated from a nuclear fuel assembly system containing nuclear fuels and outputs an alarm. The device of the present invention comprises (1) a neutron flux supply source for increasing and decreasing neutron fluxes periodically and supplying them to nuclear fuel assemblies, (2) a detector for detecting neutron fluxes in the nuclear fuel assemblies, (3) a critical state judging section for judging the critical state of the nuclear fuel assemblies based on the periodically changing signals obtained from the detector (2) and (4) an alarm section for outputting criticality alarms depending on the result of the judgement. The device of the present invention can accurately recognize the critical state of the nuclear fuel assembly system and can forecast reaching of the nuclear fuel assembly to criticality or prompt neutron critical state. (I.S.)

Study of neutron noise from reflected, metal assemblies with criticality safety applications in mind

International Nuclear Information System (INIS)

Barnett, C.S.

1985-01-01

The author studied the statistics of detected neutrons that leaked from four subcritical reflected, enriched-uranium assemblies, to explore the feasibility of developing a criticality warning system based on neutron noise analysis. The calculated multiplication factors of the assemblies are 0.59, 0.74, 0.82, and 0.92. The author studied three possible discriminators, i.e., three signatures that might be used to discriminate among assemblies of various multiplications. They are (1) variance-to-mean ratio of the counts in a time bin (V/M), (2) covariance-to-mean ratio of the counts in a common time bin from two different detectors (C/M), (3) covariance-to-mean ratio of the counts from a single detector in two adjacent time bins of equal length, which the author calls the serial-covariance-to-mean ratio (SC/M). The performances of the three discriminators were not greatly different, but a hierarchy did emerge: SC/M greater than or equal to V/M greater than or equal to C/M. An example of some results: in the neighborhood of k = 0.6 the Δk required for satisfactory discrimination varies from about 3% to 7% as detector solid angle varies from 19% to 5%. In the neighborhood of k = 0.8 the corresponding Δks are 1% and 2%. The noise analysis techniques studied performed well enough in deeply subcritical situations to deserve testing in an applications environment. They have a good chance of detecting changes in reactivity that are potentially dangerous. One can expect sharpest results when doing comparison, i.e., when comparing two records, one taken in the past under circumstances known to be normal and one taken now to search for change

Study of neutron noise from reflected, metal assemblies with criticality safety applications in mind

International Nuclear Information System (INIS)

Barnett, C.S.

1985-01-01

The author studied the statistics of detected neutrons that leaked from four subcritical reflected, enriched-uranium assemblies, to explore the feasibility of developing a criticality warning system based on neutron noise analysis. The calculated multiplication factors of the assemblies are 0.59, 0.74, 0.82, and 0.92. The author studied three possible discriminators, i.e., three signatures that might be used to discriminate among assemblies of various multiplications. They are: (1) variance-to-mean ratio of the counts in a time bin (V/M); (2) covariance-to-mean ratio of the counts in a common time bin from two different detectors (C/M); and (3) covariance-to-mean ratio of the counts from a single detector in two adjacent time bins of equal length, which the author calls the serial-covariance-to-mean ratio (SC/M). The performances of the three discriminators were not greatly different, but a hierarchy did emerge: SC/M greater than or equal to V/M greater than or equal to C/M. An example of some results: in the neighborhood of k = 0.6 the Δk required for satisfactory discrimination varies from about 3% to 7% as detector solid angle varies from 19% to 5%. In the neighborhood of k = 0.8 the corresponding Δks are 1% and 2%. The noise analysis techniques studied performed well enough in deeply subcritical situations to deserve testing in an applications environment. They have a good chance of detecting changes in reactivity that are potentially dangerous. One can expect sharpest results when doing comparisons, i.e., when comparing two records, one taken in the past under circumstances known to be normal and one taken now to search for change

Fire criticality probability analysis for 300 Area N Reactor fuel fabrication and storage facility. Revision 1

Energy Technology Data Exchange (ETDEWEB)

Kelly, J.E.

1995-02-08

Uranium fuel assemblies and other uranium associated with the shutdown N Reactor are stored in the 300 Area N Reactor Fuel Fabrication and Storage Facility (Facility). The 3712 Building, where the majority of the fuel assemblies and other uranium is stored, is where there could be a potential for a criticality bounding case. The purpose of this study is to evaluate the probability of potential fires in the Facility 3712 Building that could lead to criticality. This study has been done to support the criticality update. For criticality to occur, the wooden fuel assembly containers would have to burn such that the fuel inside would slump into a critical geometry configuration, a sufficient number of containers burn to form an infinite wide configuration, and sufficient water (about a 17 inch depth) be placed onto the slump. To obtain the appropriate geometric configuration, enough fuel assembly containers to form an infinite array on the floor would have to be stacked at least three high. Administrative controls require the stacks to be limited to two high for 1.25 wt% enriched finished fuel. This is not sufficient to allow for a critical mass regardless of the fire and accompanying water moderation. It should be noted that 0.95 wt% enriched fuel and billets are stacked higher than only two high. In this analysis, two initiating events will be considered. The first is a random fire that is hot enough and sufficiently long enough to burn away the containers and fuel separators such that the fuel can establish a critical mass. The second is a seismically induced fire with the same results.

Analysis of criticality experiments at SHE

International Nuclear Information System (INIS)

Takano, Makoto; Doi, Takeshi; Hirano, Mitsumasa; Shindo, Ryuichi; Oomura, Hiroshi

1982-03-01

In the report, the criticality experiments, which were conducted for the core configurations of Semi-Homogeneous Experimental Assembly (SHE)-8,12,13,14, are analyzed for the purpose of verifying the computer codes and calculational methods employed in the nuclear design of VHTR. The codes, DELIGHT-5 and CITATION calculate the neutron spectrum and the effective multiplication factor respectively. Each system of SHE is modeled by twodimensional R-Z, Triangular and threedimensional Triangular-Z geometries. Various effects such as axial buckling, modeling and the difference between diffusion and transport are also taken into account. Calculated values of effective multiplication factor show the disagreement of 1 - 3% from the values of experiments approximately. Therefore the analysis is considered to be inadequate to the verification and more precise analysis is required with the emphasis on how to model the system, condense the group constants and guess the buckling value for spectrum calculation. (author)

Critical residues in the PMEL/Pmel17 N-terminus direct the hierarchical assembly of melanosomal fibrils

Science.gov (United States)

Leonhardt, Ralf M.; Vigneron, Nathalie; Hee, Jia Shee; Graham, Morven; Cresswell, Peter

2013-01-01

PMEL (also called Pmel17 or gp100) is a melanocyte/melanoma-specific glycoprotein that plays a critical role in melanosome development by forming a fibrillar amyloid matrix in the organelle for melanin deposition. Although ultimately not a component of mature fibrils, the PMEL N-terminal region (NTR) is essential for their formation. By mutational analysis we establish a high-resolution map of this domain in which sequence elements and functionally critical residues are assigned. We show that the NTR functions in cis to drive the aggregation of the downstream polycystic kidney disease (PKD) domain into a melanosomal core matrix. This is essential to promote in trans the stabilization and terminal proteolytic maturation of the repeat (RPT) domain–containing MαC units, precursors of the second fibrillogenic fragment. We conclude that during melanosome biogenesis the NTR controls the hierarchical assembly of melanosomal fibrils. PMID:23389629

A Critical Appraisal of RAFT-Mediated Polymerization-Induced Self-Assembly

Science.gov (United States)

2016-01-01

Recently, polymerization-induced self-assembly (PISA) has become widely recognized as a robust and efficient route to produce block copolymer nanoparticles of controlled size, morphology, and surface chemistry. Several reviews of this field have been published since 2012, but a substantial number of new papers have been published in the last three years. In this Perspective, we provide a critical appraisal of the various advantages offered by this approach, while also pointing out some of its current drawbacks. Promising future research directions as well as remaining technical challenges and unresolved problems are briefly highlighted. PMID:27019522

WESTINGHOUSE 17X17 MOX PWR ASSEMBLY - WASTE PACKAGE CRITICALITY ANALYSIS (SCPB: N/A)

International Nuclear Information System (INIS)

J.W. Davis

1996-01-01

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to compare the criticality potential of Westinghouse 17 x 17 mixed oxide (MOX) PWR fuel with the Design Basis spent nuclear fuel (SNF) analyzed previously (Ref. 5.1, 5.2). The basis of comparison will be the conceptual design Multi-Purpose Canister (MPC) PWR waste package concepts. The objectives of this evaluation are to show that the criticality potential of the MOX fuel is equal to or lower than the DBF or, if necessary, indicate what additional measures are required to make it so

Impact of up-to-date evaluated nuclear data files on the Monte-Carlo analysis results of metallic fueled BFS critical assemblies

International Nuclear Information System (INIS)

Yoo, Jaewoon; Kim, Do-Heon; Kim, Sang-Ji; Kim, Yeong-Il

2009-01-01

Three metallic fueled BFS critical assemblies, BFS-73-1, BFS-75-1, and BFS-55-1 were analyzed by using the Monte-Carlo analysis code MCNP4C with five different evaluated data files, ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, JENDL-AC and ENDF/B-VI.6. The impacts of microscopic cross sections in the up-to-date evaluated nuclear data files were clarified by the analyses. The update of Zr cross section leads to the calculated k-effective lower than that of ENDF/B-VI.6. The revision of U-238 inelastic scattering cross section makes large difference in the predicted k-effectives between the libraries, which depends on the amount of the contribution of the inelastic cross sections change and the compensation of other reaction types. The results of the spectral indices and reaction rate ratios shows the improvement of the up-to-date evaluated nuclear data files for the U-238, Np-237, Pu-240 fission reactions, however, there are still need of further improvement for other minor actinide cross sections. The heterogeneity effects involved on the k-effective and relative fission rate distribution were evaluated in this study, which can be used as the correction factor for constructing the homogeneous benchmark configuration while keeping the consistency with the actual critical experiment. (author)

MOCA, Criticality of VVER Reactor Hexagonal Fuel Assemblies

International Nuclear Information System (INIS)

KYNCL, Jan

1994-01-01

1 - Description of program or function: Criticality problem in neutron transport for hexagonal fuel assembly in VVER nuclear reactor. The assembly is assumed to be either arranged in an infinite hexagonal array or placed in vacuum. The problem is solved in three- dimensional geometry, using standard energy group formalism and assuming that effective scattering cross sections are presented as Legendre polynomial expansions. The code evaluates ten different physical quantities, e.g. multiplication factor, neutron flux per energy group and spatial zone, integrated over angle and power in any zone of the assembly. 2 - Method of solution: Monte Carlo method of successive generations is applied. Computation proceeds according to an analog random process. The code is organized into three blocks: In the first block, the input data are converted to quantities for use in the Monte Carlo calculation. An initial neutron distribution is calculated, which corresponds to a fission spectrum uniform in spatial and angular variables. The main calculations are carried out in the second block (subroutine PROC2). This block is subdivided into geometrical and physical parts. Neutron tracks in individual zones and groups as well as probabilities for the formation of secondary neutrons are calculated. In the third block (subroutine PROC3), the results are evaluated statistically. Effective multiplication coefficients, the neutron flux per group and zone, and respective errors are computed. These quantities serve as a basis for the evaluation of other quantities. The results are either printed or stored for future evaluations. 3 - Restrictions on the complexity of the problem: In the PC version of the program, the maximum number of neutrons is 1000, the maximum number of energy groups is 4, and the maximum number of material compositions is 15. Angular expansion of scattering cross sections is allowed up to P10. These restrictions can easily be removed by increasing input parameters and

ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

Energy Technology Data Exchange (ETDEWEB)

Kahler, A.C.; Herman, M.; Kahler,A.C.; MacFarlane,R.E.; Mosteller,R.D.; Kiedrowski,B.C.; Frankle,S.C.; Chadwick,M.B.; McKnight,R.D.; Lell,R.M.; Palmiotti,G.; Hiruta,H.; Herman,M.; Arcilla,R.; Mughabghab,S.F.; Sublet,J.C.; Trkov,A.; Trumbull,T.H.; Dunn,M.

2011-12-01

The ENDF/B-VII.1 library is the latest revision to the United States Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 423 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [M. B. Chadwick et al., 'ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data,' Nuclear Data Sheets, 112, 2887 (2011)]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Project's International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unmoderated and uranium reflected {sup 235}U and {sup 239}Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also

Experience in WWER fuel assemblies vibration analysis

International Nuclear Information System (INIS)

Ovtcharov, O.; Pavelko, V.; Usanov, A.; Arkadov, G.; Dolgov, A.; Molchanov, V.

2003-01-01

It is stated that the vibration studies of internals and the fuel assemblies should be conducted during the reactor designing, commissioning and commercial operation stages and the analysis methods being used should complement each other. The present paper describes the methods and main results of the vibration noise studies of internals and the fuel assemblies of the operating NPPs with WWER reactors, as an example of the implementation of the comprehensive approach to the analysis on equipment flow-induced vibration. At that, the characteristics of internals and fuel assemblies vibration loading were dealt jointly as they are elements of the same compound oscillating system and their vibrations have the interrelated nature

Critical heat flux tests for a 12 finned-element assembly

Energy Technology Data Exchange (ETDEWEB)

Yang, J., E-mail: Jun.Yang@cnl.ca; Groeneveld, D.C.; Yuan, L.Q.

2017-03-15

Highlights: • CHF tests for a 12 finned-fuel-element assembly at highly subcooled conditions. • Test approach to maximize experimental information and minimize heater failures. • Three series of tests were completed in vertical upward light water flow. • Bundle simulators of two axial power profiles and three heated lengths were tested. • Results confirm that the prediction method predicts lower CHF values than measured. - Abstract: An experimental study was undertaken to provide relevant data to validate the current critical heat flux (CHF) prediction method of the NRU driver fuel for safety analysis, i.e., to confirm no CHF occurrence below the predicted values. The NRU driver fuel assembly consists of twelve finned fuel elements arranged in two rings – three in the inner ring and nine in the outer ring. To satisfy the experimental objective tests at very high heat fluxes, very high mass velocities, and high subcoolings were conducted where the CHF mechanism is the departure from nucleate boiling (DNB). Such a CHF experiment can be very difficult, costly and time consuming since failure of the heating surface due to rupture or melting (physical burnout) is expected when the DNB type of CHF is reached. A novel experimental approach has been developed to maximize the amount of relevant experimental information on safe operating conditions in the tests, and to minimize any possible heater failures that inherently accompany the CHF occurrence at these conditions. Three series of tests using electrically heated NRU driver fuel simulators with three heated lengths and two axial power profiles (or axial heat flux distribution (AFD)) were completed in vertical upward light water flow. Each series of tests covered two mass flow rates, several heat flux levels, and local subcoolings that bound the ranges of interest for the analysis of postulated slow loss-of-regulation accident (LORA) and loss-of-flow accident (LOFA) scenarios. Tests for each mass flow rate of

Nuclear criticality safety analysis for the traveller PWR fuel shipping package

Energy Technology Data Exchange (ETDEWEB)

Vescovi, P.J.; Kent, N.A.; Casado, C.A. [Westinghouse Electric Co., LLC, Columbia, SC (United States)]|[ENUSA Industrias Avanzadas SA, Madrid (Spain)

2004-07-01

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

Nuclear criticality safety analysis for the traveller PWR fuel shipping package

International Nuclear Information System (INIS)

Vescovi, P.J.; Kent, N.A.; Casado, C.A.

2004-01-01

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

The effect of grid assembly mixing vanes on critical heat flux values and azimuthal location in fuel assemblies

International Nuclear Information System (INIS)

De Crecy, F.

1994-01-01

Critical heat flux (CHF) is one of the limiting phenomena for a PWR. It has been widely studied for years, but many facts are still not satisfactorily understood. This paper deals with the effect of the grid assembly mixing vanes on both the value of the CHF and the azimuthal location of the departure from nucleate boiling (DNB). A series of experimental studies was performed on electrically heated, 5x5 square pitched, vertical rod bundles. Two specific grid assembly designs were used: with and without mixing vanes. DNB was detected by eight thermocouples welded internally in each rod at the same level in order to determine the azimuthal location. The coolant was Freon-12 flowing upwards to simulate high pressure water (as defined by Stevens). Single-phase flow experiments were also conducted to measure the exit temperature field in order to obtain the mixing coefficients for subchannel analysis.The results show very clearly that the mixing vanes have a significant effect on both the DNB azimuthal location and the CHF value. - Without mixing vanes, DNB occurs mainly on the most central rod and preferentially at the azimuthal location facing the adjacent rod. - With mixing vanes, DNB can occur on any of the nine central rods and is distributed in an apparently random way around the rod. -The effect of the mixing vanes on CHF is dramatic and depends a great deal on the parameter range (pressure, local mass velocity and local quality). Generally speaking, CHF with mixing vanes is significantly higher than without mixing vanes, but this effect can be inverted in some cases.In order to understand this fact more clearly, it is necessary to perform detailed analysis of subchannel behavior. Indeed, the analyses show that the magnitude of this effect is closely related to the mixing coefficients used. These mixing coefficients, estimated from the single-phase flow experiments, are subject to large uncertainties in two-phase flow. ((orig.))

The experimental determination of the buckling in the bare heavy water natural uranium critical assembly 'RB'

Energy Technology Data Exchange (ETDEWEB)

Raisic, N M; Popovic, D D; Takac, S M; Djordjevic, M M [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

1960-03-15

The buckling in the bare heavy water natural uranium critical assembly was determined by measuring the thermal neutron flux distribution. The obtained value for the critical buckling at the temperature of 20 deg C is: B{sup 2} = (8.516 {+-} 0.02) m{sup -2}. The above error is a statistical one, obtained from several series of measurements. The possible systematic error was estimated as 0.1 m{sup -2}. (author)

The experimental determination of the buckling in the bare heavy water natural uranium critical assembly 'RB'

International Nuclear Information System (INIS)

Raisic, N.M.; Popovic, D.D.; Takac, S.M.; Djordjevic, M.M.

1960-01-01

The buckling in the bare heavy water natural uranium critical assembly was determined by measuring the thermal neutron flux distribution. The obtained value for the critical buckling at the temperature of 20 deg C is: B 2 = (8.516 ± 0.02) m -2 . The above error is a statistical one, obtained from several series of measurements. The possible systematic error was estimated as 0.1 m -2 . (author)

SCALE-4 Analysis of LaSalle Unit 1 BWR Commercial Reactor Critical Configurations

International Nuclear Information System (INIS)

Gauld, I.C.

2000-01-01

Five commercial reactor criticals (CRCs) for the LaSalle Unit 1 boiling-water reactor have been analyzed using KENO V.a, the Monte Carlo criticality code of the SCALE 4 code system. The irradiated fuel assembly isotopics for the criticality analyses were provided by the Waste Package Design team at the Yucca Mountain Project in the US, who performed the depletion calculations using the SAS2H sequence of SCALE 4. The reactor critical measurements involved two beginning-of-cycle and three middle-of-cycle configurations. The CRCs involved relatively low-cycle burnups, and therefore contained a relatively high gadolinium poison content in the reactor assemblies. This report summarizes the data and methods used in analyzing the critical configurations and assesses the sensitivity of the results to some of the modeling approximations used to represent the gadolinium poison distribution within the assemblies. The KENO V.a calculations, performed using the SCALE 44GROUPNDF5 ENDF/B-V cross-section library, yield predicted k eff values within about 1% Δk/k relative to reactor measurements for the five CRCs using general 8-pin and 9-pin heterogeneous gadolinium poison pin assembly models

Solution High-Energy Burst Assembly (SHEBA) results from subprompt critical experiments with uranyl fluoride fuel

International Nuclear Information System (INIS)

Cappiello, C.C.; Butterfield, K.B.; Sanchez, R.G.; Bounds, J.A.; Kimpland, R.H.; Damjanovich, R.P.; Jaegers, P.J.

1997-01-01

Experiments were performed to measure a variety of parameters for SHEBA: behavior of the facility during transient and steady-state operation; characteristics of the SHEBA fuel; delayed-critical solution height vs solution temperature; initial reactor period and reactivity vs solution height; calibration of power level vs reactor power instrumentation readings; flux profile in SHEBA; radiation levels and neutron spectra outside the assembly for code verification and criticality alarm and dosimetry purposes; and effect on reactivity of voids in the fuel

Evaluation of neutron flux in the Pool Critical Assembly

International Nuclear Information System (INIS)

Lippincott, E.P.; Ruddy, F.H.; Gold, R.; Kellogg, L.S.; Roberts, J.H.

1984-09-01

A recently completed series of experiments in the Pool Critical Assembly (PCA) at Oak Ridge National Laboratory (ORNL) provided extensive neutron flux characterization of a mockup pressure vessel configuration. Considerable effort has been made to understand the uncertainties of the various measurements made in the PCA and to resolve discrepancies in the data. Additional measurements are available for similar configurations in the Oak Ridge Reactor-Poolside Facility (ORR-PSF) at ORNL and in the NESDIP facility in the UK. Comparisons of these results, together with associated neutron field calculations, enable a better evaluation of the actual uncertainties and realistic limits of accuracy to be assessed. Such assessments are especially valuable when the accuracy improvements of benchmark referencing are to be included and extrapolations to new configurations are made

Measurement of the neutron spectrum of the Big Ten critical assembly by lithium-6 spectrometry

International Nuclear Information System (INIS)

De Leeuw-Gierts, G.; De Leeuw, S.; Hansen, G.E.; Helmick, H.H.

1979-01-01

The central neutron-flux spectrum of the Los Alamos Scientific Laboratory's critical assembly, Big Ten, was measured with a 6 Li spectrometer and techniques developed at the Centre d'Etude de L'Energie Nucleaire, Mol, as part of an experimental program to establish the characteristics of Big Ten

Measurement of the neutron spectrum of the Big Ten critical assembly by lithium-6 spectrometry

International Nuclear Information System (INIS)

Leeuw-Gierts, G. de; Leeuw, S. de

1980-01-01

The central neutron-flux spectrum of the Los Alamos Scientific Laboratory's critical assembly, Big Ten, was measured with a 6 Li spectrometer and techniques developed at the Centre d'Etude de l'Energie Nucleaire, Mol, as part of an experimental program to establish the characteristics of Big Ten

Monte Carlo analysis of Pu-H2O and UO2-PuO2-H2O critical assemblies with ENDF/B-IV data

International Nuclear Information System (INIS)

Hardy, J. Jr.; Ullo, J.J.

1981-04-01

A set of critical experiments, comprising thirteen homogeneous Pu-H 2 O assemblies and twelve UO 2 -PuO 2 lattices, was analyzed with ENDF/B-IV data and the RCPO1 Monte Carlo program, which modeled the experiments explicitly. Some major data sensitivities were also evaluated. For the Pu-H 2 O assemblies, calculated K/sub eff/ averaged 1.011. The large (2.7%) scatter of K/sub eff/ values for these assemblies was attributed mostly to uncertainties in physical specifications since no clear trends of K/sub eff/ were evident and data sensitivities were insignificant. The UO 2 -PuO 2 lattices showed just one trend of K/sub eff/, which indicated an overprediction of U238 capture consistent with that observed for uranium-H 2 O experiments. There was however a approx. 1% discrepancy in calculated K/sub eff/ between the two sets of UO 2 -PuO 2 lattices studied

Study on neutron streaming effect in large fast critical assembly

International Nuclear Information System (INIS)

Takeda, Toshikazu; Yamaoka, Mitsuaki; Sakurai, Shungo; Tanimoto, Koichi; Abe, Yuhei

1981-03-01

A cell calculation method taking into account the neutron leakage from a cell and a transport calculation method treating the neutron streaming have been developed, and their applicability has been investigated. In the cell calculation method, the neutron leakage in the perpendicular direction to plates was treated by introducing an albedo collision probability which is a first-flight collision probability incorporating albedos at cell boundaries, and that in the parallel direction was treated by the pseudo absorption method. The use of the albedo collision probability made it possible to calculate the flux tilt in a cell exactly. This cell calculation method was applied to two slab models where fuel drawers were stacked in perpendicular and parallel directions to plates. Cell averaged cross sections calculated by the proposed method agreed well with those obtained from exact transport calculations treating the plate-wise heterogeneity, while the infinite cell calculation and the conventional pseudo absorption method produced about 2% errors in the cell-averaged cross sections. The cell-averaging procedure for control-rod channels was also proposed, and this method was applied to the calculation of control-rod worths and control-rod position worths. A transport calculation method based on the response matrix method has been proposed to treat the neutron streaming in fast critical assemblies directly. A response matrix code in two dimensional XY geometry RES2D was made. The accuracy of response matrices obtained from the RES2D code was checked by applying it to a slab cell and by comparing cell-averaged cross sections and k-infinity with those from a reference cell calculation based on the collision probability. The agreement of the results was good, and it was found that the response matrix method is very promising for the treatment of the neutron streaming in fast critical assemblies. (author)

Multicriteria Analysis of Assembling Buildings from Steel Frame Structures

Science.gov (United States)

Miniotaite, Ruta

2017-10-01

Steel frame structures are often used in the construction of public and industrial buildings. They are used for: all types of slope roofs; walls of newly-built public and industrial buildings; load bearing structures; roofs of renovated buildings. The process of assembling buildings from steel frame structures should be analysed as an integrated process influenced by such factors as construction materials and machinery used, the qualification level of construction workers, complexity of work, available finance. It is necessary to find a rational technological design solution for assembling buildings from steel frame structures by conducting a multiple criteria analysis. The analysis provides a possibility to evaluate the engineering considerations and find unequivocal solutions. The rational alternative of a complex process of assembling buildings from steel frame structures was found through multiple criteria analysis and multiple criteria evaluation. In multiple criteria evaluation of technological solutions for assembling buildings from steel frame structures by pairwise comparison method the criteria by significance are distributed as follows: durability is the most important criterion in the evaluation of alternatives; the price (EUR/unit of measurement) of a part of assembly process; construction workers’ qualification level (category); mechanization level of a part of assembling process (%), and complexity of assembling work (in points) are less important criteria.

A Study on the Fuel Assembly Seismic Analysis without Holddown Springs

International Nuclear Information System (INIS)

Kwon, O Cheol; Ha, Dong Geun; Lee, Kyou Seok; Jeon, Sang Yoon; Suh, Jung Min

2013-01-01

In this study, the effect for the fuel assembly removed holddown spring under seismic event has been evaluated through the comparison with the seismic analysis result of fuel assembly with holddown spring. In order to compare each design, the simplified fuel assembly seismic analysis models have been established according to reference. The mid grid impact force, natural frequency, and top nozzle displacement for each fuel assembly model has been analyzed using ANSYS. The fuel assembly seismic analyses without holddown springs are performed and compared to the model with holddown springs. The grid impact forces of CPM 1 and CPM 2 are almost doubled in comparison with CPM 3 and almost tripled in comparison with CPM 4 so the grid impact forces depend on CPM types. The grid impact forces of the fuel assembly model without holddown springs have similar tendencies in comparison with fuel assembly with holddown springs. Moreover, the model without holddown springs analysis time is much longer than the model with holddown springs. Consequently, it is moderate that the fuel assembly analysis model with holddown springs would be used for effective analysis even though the actual model has no holddown springs

A Study on the Fuel Assembly Seismic Analysis without Holddown Springs

Energy Technology Data Exchange (ETDEWEB)

Kwon, O Cheol; Ha, Dong Geun; Lee, Kyou Seok; Jeon, Sang Yoon; Suh, Jung Min [KEPCO Nuclear Fuel, Daejeon (Korea, Republic of)

2013-10-15

In this study, the effect for the fuel assembly removed holddown spring under seismic event has been evaluated through the comparison with the seismic analysis result of fuel assembly with holddown spring. In order to compare each design, the simplified fuel assembly seismic analysis models have been established according to reference. The mid grid impact force, natural frequency, and top nozzle displacement for each fuel assembly model has been analyzed using ANSYS. The fuel assembly seismic analyses without holddown springs are performed and compared to the model with holddown springs. The grid impact forces of CPM{sub 1} and CPM{sub 2} are almost doubled in comparison with CPM{sub 3} and almost tripled in comparison with CPM{sub 4} so the grid impact forces depend on CPM types. The grid impact forces of the fuel assembly model without holddown springs have similar tendencies in comparison with fuel assembly with holddown springs. Moreover, the model without holddown springs analysis time is much longer than the model with holddown springs. Consequently, it is moderate that the fuel assembly analysis model with holddown springs would be used for effective analysis even though the actual model has no holddown springs.

Numerical verification of the theory of coupled reactors for a deuterium critical assembly using MCNP5

International Nuclear Information System (INIS)

Hussein, M.S.; Bonin, H.W.; Lewis, B.J.

2013-01-01

The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as Deuterium Critical Assembly, (DCA). The variations of the criticality factors and the coupling coefficients were investigated by changing of the water levels in the inner and outer cores. The numerical results of the model developed with MCNP5 code were validated and verified against published results and the mathematical model based on coupled reactor theory. (author)

Numerical verification of the theory of coupled reactors for a deuterium critical assembly using MCNP5

Energy Technology Data Exchange (ETDEWEB)

Hussein, M.S.; Bonin, H.W.; Lewis, B.J., E-mail: mohamed.hussein@rmc.ca, E-mail: bonin-h@rmc.ca, E-mail: lewis-b@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

2013-07-01

The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as Deuterium Critical Assembly, (DCA). The variations of the criticality factors and the coupling coefficients were investigated by changing of the water levels in the inner and outer cores. The numerical results of the model developed with MCNP5 code were validated and verified against published results and the mathematical model based on coupled reactor theory. (author)

Monte Carlo cross section testing for thermal and intermediate 235U/238U critical assemblies, ENDF/B-V vs ENDF/B-VI

International Nuclear Information System (INIS)

Weinman, J.P.

1997-06-01

The purpose of this study is to investigate the eigenvalue sensitivity to changes in ENDF/B-V and ENDF/B-VI cross section data sets by comparing RACER vectorized Monte Carlo calculations for several thermal and intermediate spectrum critical experiments. Nineteen Oak Ridge and Rocky Flats thermal solution benchmark critical assemblies that span a range of hydrogen-to- 235 U (H/U) concentrations (2052 to 27.1) and above-thermal neutron leakage fractions (0.555 to 0.011) were analyzed. In addition, three intermediate spectrum critical assemblies (UH3-UR, UH3-NI, and HISS-HUG) were studied

A spent fuel assemblies monitoring device by nondestructive analysis 'PYTHON'

International Nuclear Information System (INIS)

Saad, M.; Broeskamp, M.; Hahn, H.; Bignan, G.; Boisset, M.; Silie, P.

1995-01-01

The monitoring of spent fuel assemblies (16 x 16 UOX) in KWG-reactor pool with the use of non-destructive methods (total Gamma and neutron counting) allow the control of average burn-up and the extremity burn-up. The measurements allow a safety-criticality control before loading the fuel assemblies into the transport casks. A device called PYTHON has been tested and qualified in France. This paper presents a description of the industrial PYTHON device and the results of the measurements. (orig.)

Characterization of neutron leakage probability, k /SUB eff/ , and critical core surface mass density of small reactor assemblies through the Trombay criticality formula

International Nuclear Information System (INIS)

Kumar, A.; Rao, K.S.; Srinivasan, M.

1983-01-01

The Trombay criticality formula (TCF) has been derived by incorporating a number of well-known concepts of criticality physics to enable prediction of changes in critical size or k /SUB eff/ following alterations in geometrical and physical parameters of uniformly reflected small reactor assemblies characterized by large neutron leakage from the core. The variant parameters considered are size, shape, density and diluent concentration of the core, and density and thickness of the reflector. The effect of these changes (except core size) manifests, through sigma /SUB c/ the critical surface mass density of the ''corresponding critical core,'' that sigma, the massto-surface-area ratio of the core,'' is essentially a measure of the product /rho/ extended to nonspherical systems and plays a dominant role in the TCF. The functional dependence of k /SUB eff/ on sigma/sigma /SUB c/ , the system size relative to critical, is expressed in the TCF through two alternative representations, namely the modified Wigner rational form and, an exponential form, which is given

Solution High-Energy Burst Assembly (SHEBA) results from subprompt critical experiments with uranyl fluoride fuel

International Nuclear Information System (INIS)

Cappiello, C.C.; Butterfield, K.B.; Sanchez, R.G.

1997-10-01

The Solution High-Energy Burst Assembly (SHEBA) was originally constructed during 1980 and was designed to be a clean free-field geometry, right-circular, cylindrically symmetric critical assembly employing U(5%)O 2 F 2 solution as fuel. A second version of SHEBA, employing the same fuel but equipped with a fuel pump and shielding pit, was commissioned in 1993. This report includes data and operating experience for the 1993 SHEBA only. Solution-fueled benchmark work focused on the development of experimental measurements of the characterization of SHEBA; a summary of the results are given. A description of the system and the experimental results are given in some detail in the report. Experiments were designed to: (1) study the behavior of nuclear excursions in a low-enrichment solution, (2) evaluate accidental criticality alarm detectors for fuel-processing facilities, (3) provide radiation spectra and dose measurements to benchmark radiation transport calculations on a low-enrichment solution system similar to centrifuge enrichment plants, and (4) provide radiation fields to calibrate personnel dosimetry. 15 refs., 37 figs., 10 tabs

Neutron data testing for plutonium isotopes in experiments at fast critical assemblies

International Nuclear Information System (INIS)

Bednyakov, S.M.; Dulin, V.A.; Manturov, G.N.; Mozhaev, V.K.; Semenov, M.Yu.; Tsibulya, A.M.

1996-01-01

Experimental results on checking neutron data, obtained at the fast critical assemblies, are presented. They constitute sufficiently large collection of data making it possible to test nuclear neutron constants of plutonium isotopes for the new system of group constants BNAB-93. The work contains comparison of the measurement results on average fission cross section ratios and reactivity coefficients ratios for 239,240,241 Pu (to 235 U) with calculational data, obtained on the basis of the new testing system of the BNAB-93 group constants system. 14 refs., 6 figs

Numerical verification/validation of the theory of coupled reactors for deuterium critical assembly, using MCNP5 and Serpent codes

International Nuclear Information System (INIS)

Hussein, M.S; Lewis, B.J.; Bonin, H.W.

2013-01-01

The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5 and the continuous-energy Monte Carlo reactor physics burnup calculation Serpent code. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as the Deuterium Critical Assembly, DCA. The multiplication factors k eff calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors k eff calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors k eff calculated with the Serpent code, with MCNP5, and from the coupled reactor theory. This analysis demonstrates that the Serpent code is valid for the multipoint coupled reactor calculations. (author)

Numerical verification/validation of the theory of coupled reactors for deuterium critical assembly, using MCNP5 and Serpent codes

Energy Technology Data Exchange (ETDEWEB)

Hussein, M.S, E-mail: mohamed.hussein@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada); Lewis, B.J., E-mail: Brent.Lewis@uoit.ca [Univ. of Ontario Inst. of Technology, Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada); Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

2013-07-01

The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5 and the continuous-energy Monte Carlo reactor physics burnup calculation Serpent code. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as the Deuterium Critical Assembly, DCA. The multiplication factors k{sub eff} calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors k{sub eff} calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors k{sub eff} calculated with the Serpent code, with MCNP5, and from the coupled reactor theory. This analysis demonstrates that the Serpent code is valid for the multipoint coupled reactor calculations. (author)

Sensitivity Analysis of Deviation Source for Fast Assembly Precision Optimization

Directory of Open Access Journals (Sweden)

Jianjun Tang

2014-01-01

Full Text Available Assembly precision optimization of complex product has a huge benefit in improving the quality of our products. Due to the impact of a variety of deviation source coupling phenomena, the goal of assembly precision optimization is difficult to be confirmed accurately. In order to achieve optimization of assembly precision accurately and rapidly, sensitivity analysis of deviation source is proposed. First, deviation source sensitivity is defined as the ratio of assembly dimension variation and deviation source dimension variation. Second, according to assembly constraint relations, assembly sequences and locating, deviation transmission paths are established by locating the joints between the adjacent parts, and establishing each part’s datum reference frame. Third, assembly multidimensional vector loops are created using deviation transmission paths, and the corresponding scalar equations of each dimension are established. Then, assembly deviation source sensitivity is calculated by using a first-order Taylor expansion and matrix transformation method. Finally, taking assembly precision optimization of wing flap rocker as an example, the effectiveness and efficiency of the deviation source sensitivity analysis method are verified.

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

International Nuclear Information System (INIS)

Wilson, Michael L.

2001-01-01

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

Characterization of the CALIBAN Critical Assembly Neutron Spectra using Several Adjustment Methods Based on Activation Foils Measurement

OpenAIRE

Casoli Pierre; Grégoire Gilles; Rousseau Guillaume; Jacquet Xavier; Authier Nicolas

2016-01-01

CALIBAN is a metallic critical assembly managed by the Criticality, Neutron Science and Measurement Department located on the French CEA Center of Valduc. The reactor is extensively used for benchmark experiments dedicated to the evaluation of nuclear data, for electronic hardening or to study the effect of the neutrons on various materials. Therefore CALIBAN irradiation characteristics and especially its central cavity neutron spectrum have to be very accurately evaluated. In order to streng...

Critical eigenvalue in LMFBRs: a physics assessment

International Nuclear Information System (INIS)

McKnight, R.D.; Collins, P.J.; Olsen, D.N.

1984-01-01

This paper summarizes recent work to put the analysis of past critical eigenvalue measurements from the US critical experiments program on a consistent basis. The integral data base includes 53 configurations built in 11 ZPPR assemblies which simulate mixed oxide LMFBRs. Both conventional and heterogeneous designs representing 350, 700, and 900 MWe sizes and with and without simulated control rods and/or control rod positions have been studied. The review of the integral data base includes quantitative assessment of experimental uncertainties in the measured excess reactivity. Analyses have been done with design level and higher-order methods using ENDF/B-IV data. Comparisons of these analyses with the experiments are used to generate recommended bias factors for criticality predictions. Recommended methods for analysis of LMFBR fast critical assemblies and LMFBR design calculations are presented. Unresolved issues and areas which require additional experimental or analytical study are identified

Examination of Successful Modal Analysis Techniques Used for Bladed-Disk Assemblies

National Research Council Canada - National Science Library

Orsagh, R

2002-01-01

Modal testing of bladed-disk assemblies in turbomachines is used to identify the critical natural frequencies and mode shape information used for avoiding the per-rev resonant conditions that cause high cycle fatigue (HCF...

Studies of spatial decoupling in heterogeneous LMFBR critical assemblies

International Nuclear Information System (INIS)

Brumbach, S.B.; Goin, R.W.; Carpenter, S.G.

1984-01-01

Recent measurements at the Zero Power Plutonium Reactor have studied the spatial decoupling in large, heterogeneous assemblies. These assemblies exhibited a significantly greater degree of decoupling than previous homogeneous assemblies of similar size. The flux distributions in these heterogeneous assemblies were very sensitive reactivity perturbations, and perturbed flux distributions were achieved relatively slowly. Decoupling was investigated using rod-drop, boron-oscillator and noise-coherence techniques which emphasized different times following the perturbations. Reactivity changes could be measured by analyzing the power history from a single detector using inverse kinetics methods with the assumption of an instantaneous efficiency change for the detector. For assemblies more decoupled than ZPPR-13, the instantaneous efficiency change assumption begins to be invalid

Colloidal Self-Assembly Driven by Deformability & Near-Critical Phenomena

NARCIS (Netherlands)

Evers, C.H.J.|info:eu-repo/dai/nl/338775188

2016-01-01

Self-assembly is the spontaneous formation of patterns or structures without human intervention. This thesis aims to increase our understanding of self-assembly. In self-assembly of proteins, the building blocks are very small and complex. Consequently, grasping the basic principles that drive the

Method of preventing criticality of fresh fuel assembly in storage facility

International Nuclear Information System (INIS)

Kawamura, Makoto.

1990-01-01

With an aim of improving the operation efficiency of a reactor, extention of the operation cycle by increasing U 235 enrichment degree of fuel uranium is planned. However, along with the increase of the enrichment degree of the fuel uranium, there occurs a problem of criticality upon fuel handling. Then, in the present invention, boric acid incorporating B-10 of great neutron absorption effect are packed with water soluble polymeric materials which are further packed with a fuel packing sheet, or the water soluble polymeric materials incorporating boric acids are packed with fuel packing sheets which are disposed to a fresh fuel assembly and stored in a store house as they are. The fuel packing sheet is a perforated sheet having a plurality of water intruding pores. Then, if water should intrude to the store house accidentally, the water soluble polymeric materials are dissolved, so that the intruded water is converted into aqueous boric acid easily and absorbs neutrons effectively to thereby attain the prevention of criticality. (T.M.)

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

International Nuclear Information System (INIS)

Wright, Kenneth D.

1997-01-01

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

Autoradiographic technique for rapid inventory of plutonium-containing fast critical assembly fuel

International Nuclear Information System (INIS)

Brumbach, S.B.; Perry, R.B.

1977-10-01

A nondestructive autoradiographic technique is described which can provide a verification of the piece count and the plutonium content of plutonium-containing fuel elements. This technique uses the spontaneously emitted gamma rays from plutonium to form images of fuel elements on photographic film. Autoradiography has the advantage of providing an inventory verification without the opening of containers or the handling of fuel elements. Missing fuel elements, substitution of nonradioactive material, and substitution of elements of different size are detectable. Results are presented for fuel elements in various storage configurations and for fuel elements contained in a fast critical assembly

Critical Analysis of Multimodal Discourse

DEFF Research Database (Denmark)

van Leeuwen, Theo

2013-01-01

This is an encyclopaedia article which defines the fields of critical discourse analysis and multimodality studies, argues that within critical discourse analysis more attention should be paid to multimodality, and within multimodality to critical analysis, and ends reviewing a few examples of re...

Thermal structural analysis of SST-1 vacuum vessel and cryostat assembly using ANSYS

International Nuclear Information System (INIS)

Santra, Prosenjit; Bedakihale, Vijay; Ranganath, Tata

2009-01-01

Steady state super-conducting tokamak-1 (SST-1) is a medium sized tokamak, which has been designed to produce a 'D' shaped double null divertor plasma and operate in quasi steady state (1000 s). SST-1 vacuum system comprises of plasma chamber (vacuum vessel, interconnecting rings, baking and cooling channels), and cryostat all made of SS 304L material designed to meet ultra high vacuum requirements for plasma generation and confinement. Prior to plasma shot and operation the vessel assembly is baked to 250/150 deg. C from room temperature and discharge cleaned to remove impurities/trapped gases from wall surfaces. Due to baking the non-uniform temperature pattern on the vessel assembly coupled with atmospheric pressure loading and self-weight give rise to high thermal-structural stresses, which needs to be analyzed in detail. In addition the vessel assembly being a thin shell vessel structure needs to be checked for critical buckling load caused by atmospheric and baking thermal loads. Considering symmetry of SST-1, 1/16th of the geometry is modeled for finite element (FE) analysis using ANSYS for different loading scenarios, e.g. self-weight, pressure loading considering normal operating conditions, and off-normal loads coupled with baking of vacuum vessel from room temperature 250 deg. C to 150 deg. C, buckling and modal analysis for future dynamic analysis. The paper will discuss details about SST-1 vacuum system/cryostat, solid and FE model of SST-1, different loading scenarios, material details and the stress codes used. We will also present the thermal structural results of FE analysis using ANSYS for various load cases being investigated and our observations under different loading conditions.

Educational use of research reactor (KUR) and critical assembly (KUCA) at Kyoto University

International Nuclear Information System (INIS)

Misawa, Tsuyoshi; Unesaki, Hironobu; Ichihara, Chihiro; Pyeon, Cheol Ho; Shiroya, Seiji

2005-01-01

At Kyoto University Research Reactor Institute, a research reactor of 5MW (KUR) and a critical assembly (KUCA) have been used for educational purpose to train undergraduate or graduate students. Using KUR, basic experiments for neutron applications have been carried out, and KUCA has been used for the education of nuclear engineering and technology. Especially, using KUCA, a joint reactor laboratory course of graduate level is offered every summer since 1975 by nine associated Japanese universities, and more than 2200 students attended this course

Critical factors for assembling a high volume of DNA barcodes

Science.gov (United States)

Hajibabaei, Mehrdad; deWaard, Jeremy R; Ivanova, Natalia V; Ratnasingham, Sujeevan; Dooh, Robert T; Kirk, Stephanie L; Mackie, Paula M; Hebert, Paul D.N

2005-01-01

Large-scale DNA barcoding projects are now moving toward activation while the creation of a comprehensive barcode library for eukaryotes will ultimately require the acquisition of some 100 million barcodes. To satisfy this need, analytical facilities must adopt protocols that can support the rapid, cost-effective assembly of barcodes. In this paper we discuss the prospects for establishing high volume DNA barcoding facilities by evaluating key steps in the analytical chain from specimens to barcodes. Alliances with members of the taxonomic community represent the most effective strategy for provisioning the analytical chain with specimens. The optimal protocols for DNA extraction and subsequent PCR amplification of the barcode region depend strongly on their condition, but production targets of 100K barcode records per year are now feasible for facilities working with compliant specimens. The analysis of museum collections is currently challenging, but PCR cocktails that combine polymerases with repair enzyme(s) promise future success. Barcode analysis is already a cost-effective option for species identification in some situations and this will increasingly be the case as reference libraries are assembled and analytical protocols are simplified. PMID:16214753

Failure analysis of top nozzle holddown spring screw for nuclear fuel assembly

International Nuclear Information System (INIS)

Koh, S. K.; Ryu, C. H.; Na, E. G.; Baek, T. H.; Jeon, K. L.

2003-01-01

A failure analysis of holddown spring screw was performed using fracture mechanics approach. The spring screw was designed such that it was capable of sustaining the loads imposed by the initial tensile preload and operational loads. In order to investigate the cause of failure, a stress analysis of the top nozzle spring assembly was done using finite element analysis and a life prediction of the screw was made using a fracture mechanics approach. The elastic-plastic finite element analysis showed that the local stresses at the critical regions of head-shank fillet and thread root significantly exceeded than the yield strength of the screw material, resulting in local plastic deformation. Primary water stress corrosion cracking life of the Inconel 600 screw was predicted by using integration of the Scott model and resulted in 1.42 years, which was fairly close to the actual service life of the holddown spring screw

Impact Analysis for Fuel Assemblies in Spent Fuel Storage Rack

International Nuclear Information System (INIS)

Oh, Jinho

2013-01-01

The design and structural integrity evaluation of a spent fuel storage rack (SFSR) utilized for storing and protecting the spent fuel assemblies generated during the operation of a reactor are very important in terms of nuclear safety and waste management. The objective of this study is to show the validity of the SFSR design as well as fuel assembly through a structural integrity evaluation based on a numerical analysis. In particular, a dynamic time history analysis considering the gaps between the fuel assemblies and the walls of the storage cell pipes in the SFSR was performed to check the structural integrity of the fuel assembly and storage cell pipe

Impact Analysis for Fuel Assemblies in Spent Fuel Storage Rack

Energy Technology Data Exchange (ETDEWEB)

Oh, Jinho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-07-01

The design and structural integrity evaluation of a spent fuel storage rack (SFSR) utilized for storing and protecting the spent fuel assemblies generated during the operation of a reactor are very important in terms of nuclear safety and waste management. The objective of this study is to show the validity of the SFSR design as well as fuel assembly through a structural integrity evaluation based on a numerical analysis. In particular, a dynamic time history analysis considering the gaps between the fuel assemblies and the walls of the storage cell pipes in the SFSR was performed to check the structural integrity of the fuel assembly and storage cell pipe.

Safety analysis of the Los Alamos critical experiments facility

International Nuclear Information System (INIS)

Paxton, H.C.

1975-10-01

The safety of Pajarito Site critical assembly operations depends upon protection built into the facility, upon knowledgeable personnel, and upon good practice as defined by operating procedures and experimental plans. Distance, supplemented by shielding in some cases, would protect personnel against an extreme accident generating 10 19 fissions. During the facility's 28-year history, the direct cost of criticality accidents has translated to a risk of less than $200 per year

Preliminary safety analysis of criticality for dual-purpose metal cask under dry storage conditions in South Korea

Energy Technology Data Exchange (ETDEWEB)

Kim, Taeman, E-mail: tmkim@korad.or.kr [Korea Radioactive Waste Agency (KORAD), 1045 Daedeokdaero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Dho, Hoseog; Baeg, Chang-Yeal [Korea Radioactive Waste Agency (KORAD), 1045 Daedeokdaero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Lee, Gang-uk [Korea Nuclear Engineering and Service Co. (KONES), Hyundai Plaza, 341-4 Jangdae-dong, Yuseong-gu, Daejeon (Korea, Republic of)

2014-10-15

Highlights: • DPC is under development led by Korea Radioactive Waste Agency in South Korea. • The results of criticality analysis with respect to design requirements. • The k{sub eff} under normal and off-normal conditions were 0.36 and 0.46, respectively. • In addition, the k{sub eff} under a postulated accident condition was evaluated to be 0.94. - Abstract: A dual-purpose metal cask is under development led by Korea Radioactive Waste Agency (KORAD) in Korea, for the dry interim storage and long-distance transportation. This cask comprises a main body made of carbon steel and a stainless steel Dry Shielded Canister (DSC), with stainless steel baskets inside to contain spent fuel assemblies. In this study, nuclear criticality safety analysis was conducted as a part of safety assessment of the metal cask. Analysis to show criticality safety in accordance with regulatory requirements of PWR spent fuel storage was carried out. 10CFR72.124 “Criteria for nuclear criticality safety” and the Regulatory Guide of the American Nuclear Society, ANSI/ANS-57.9 “Design Criteria for an Independent Spent Fuel” and US NRC's “Standard Review Plan for Spent Fuel Dry Storage Systems at a General License Facility” were employed as regulatory standard and criteria. This paper shows results of criticality analysis with respect to each designated criterion with modeling of a virtual nuclear fuel assembly and a cask body that induces the maximum reactivity among various design basis fuels of the metal cask. In addition, the sensitivity analysis of nuclear criticality taking into account the various modeling deviation such as manufacturing tolerance and modeling assumptions of conventional models was carried out to ensure the reliability of the analysis result. The criticality evaluation result of the metal cask and the maximum k{sub eff} under normal and off-normal conditions were 0.36884 and 0.46255, respectively. The maximum k{sub eff} under a postulated

Criticality safety evaluation for TWR-S fuel assembly transportation using TK-S16 containers

International Nuclear Information System (INIS)

Pesic, M.P.; Steljic, M.M.; Antic, D.P.

2002-01-01

Criticality safety issues, concerning transportation of fresh high-enriched uranium fuel elements (TWR-S fuel assembly type) with Russian containers TK-S16, are objects of study in this paper. Three-dimensional (3D) models of fuel element and container were made, based upon their well-known geometry and material structure. The way to pack fuel elements in a bundle inside of the container is proposed. Calculations were done by MCNP4B2 computer code. This Monte Carlo criticality code determined the effective multiplication factor from the cross-section data and specific geometry data. This evaluation demonstrated the subcriticality of a single package and an array of packages during normal conditions of transport and various hypothetical accident conditions. (author)

SNEAK assembly 12B - Construction and experimental program

International Nuclear Information System (INIS)

Helm, F.

1982-10-01

The assembly SNEAK 12A is the second SNEAK assembly in which the neutron physics of disturbed material configurations, as they might occur in a postulated fast reactor accident, is investigated. The core is characterized by a central zone with fuel rods containing PuO 2 UO 2 mixed oxide pellets. This zone is cylindrically surrounded by a buffer zone for spectral adaptation, a driver zone sufficiently reactive to maintain criticality and a radial depleted uranium blanket. The purpose of the experiments is to test calculational methods of neutron physics which are used in the accident analysis of fast breeder reactors. The report describes the construction of the core and its assemblies and their content and outlines the foreseen experimental program. It includes measurements in the clean critical core and investigations of vertical and horizontal fuel relocations and steel blockages

Nanopore sequencing technology and tools for genome assembly: computational analysis of the current state, bottlenecks and future directions.

Science.gov (United States)

Senol Cali, Damla; Kim, Jeremie S; Ghose, Saugata; Alkan, Can; Mutlu, Onur

2018-04-02

Nanopore sequencing technology has the potential to render other sequencing technologies obsolete with its ability to generate long reads and provide portability. However, high error rates of the technology pose a challenge while generating accurate genome assemblies. The tools used for nanopore sequence analysis are of critical importance, as they should overcome the high error rates of the technology. Our goal in this work is to comprehensively analyze current publicly available tools for nanopore sequence analysis to understand their advantages, disadvantages and performance bottlenecks. It is important to understand where the current tools do not perform well to develop better tools. To this end, we (1) analyze the multiple steps and the associated tools in the genome assembly pipeline using nanopore sequence data, and (2) provide guidelines for determining the appropriate tools for each step. Based on our analyses, we make four key observations: (1) the choice of the tool for basecalling plays a critical role in overcoming the high error rates of nanopore sequencing technology. (2) Read-to-read overlap finding tools, GraphMap and Minimap, perform similarly in terms of accuracy. However, Minimap has a lower memory usage, and it is faster than GraphMap. (3) There is a trade-off between accuracy and performance when deciding on the appropriate tool for the assembly step. The fast but less accurate assembler Miniasm can be used for quick initial assembly, and further polishing can be applied on top of it to increase the accuracy, which leads to faster overall assembly. (4) The state-of-the-art polishing tool, Racon, generates high-quality consensus sequences while providing a significant speedup over another polishing tool, Nanopolish. We analyze various combinations of different tools and expose the trade-offs between accuracy, performance, memory usage and scalability. We conclude that our observations can guide researchers and practitioners in making conscious

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

International Nuclear Information System (INIS)

Wright, Kenneth D.

1997-01-01

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

Characterization of the Caliban and Prospero Critical Assemblies Neutron Spectra for Integral Measurements Experiments

Science.gov (United States)

Casoli, P.; Authier, N.; Jacquet, X.; Cartier, J.

2014-04-01

Caliban and Prospero are two highly enriched uranium metallic core reactors operated on the CEA Center of Valduc. These critical assemblies are suitable for integral experiments, such as fission yields measurements or perturbation measurements, which have been carried out recently on the Caliban reactor. Different unfolding methods, based on activation foils and fission chambers measurements, are used to characterize the reactor spectra and especially the Caliban spectrum, which is very close to a pure fission spectrum.

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

Energy Technology Data Exchange (ETDEWEB)

Kenneth D. Wright

1997-07-30

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

Preliminary Failure Modes, Effects and Criticality Analysis (FMECA) of the conceptual Brayton Isotope Power System (BIPS) Flight System

International Nuclear Information System (INIS)

Miller, L.G.

1976-01-01

A failure modes, effects and criticality analysis (FMECA) was made of the Brayton Isotope Power System Flight System (BIPS-FS) as presently conceived. The components analyzed include: Mini-BRU; Heat Source Assembly (HSA); Mini-Brayton Recuperator (MBR); Space Radiator; Ducts and Bellows, Insulation System; Controls; and Isotope Heat Source (IHS)

Passive gamma analysis of the boiling-water-reactor assemblies

Energy Technology Data Exchange (ETDEWEB)

Vo, D., E-mail: ducvo@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Atomic Energy Community (EURATOM), Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Atomic Energy Community (EURATOM), Luxemburg (Luxembourg)

2016-09-11

This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative–Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: {sup 137}Cs, {sup 154}Eu, {sup 134}Cs, and to a lesser extent, {sup 106}Ru and {sup 144}Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

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

Energy Technology Data Exchange (ETDEWEB)

Kenneth D. Wright

1997-09-03

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

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

International Nuclear Information System (INIS)

Wright, Kenneth D.

1997-01-01

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

Thermal hydraulic analysis of Pb-Bi cooled HYPER fuel assemblies using SLTHEN code

International Nuclear Information System (INIS)

Tak, Nam Il; Song, Tae Y.; Park, Won S.; Kim, Chang Hyun

2002-12-01

In the present work, the existing SLTHEN code, which had been originally developed for subchannel analysis of sodium cooled fast reactors, was modified and applied to the Pb-Bi cooled HYPER core which consists of 237 fuel assemblies (TRU assemblies). In the analysis of single fuel assembly having chopped cosine power profile, the validation and the assessment of usefulness of the modified SLTHEN were focused. In the quantitative comparison, the results of the modified SLTHEN agreed well with those of analytical calculations and of MATRA. For the qualitative approaches, the sensitivity calculations for intra-assembly gap flow and turbulent mixing parameter were used. The sensitivity analysis results showed that the modified SLTHEN can provide reasonable simulations of subchannel thermal hydraulics. In particular, turbulent mixing parameter which is known as the most uncertain parameter in subchannel analyses did not affect largely the maximum cladding temperature. Therefore, it can be said that the results of single assembly show the usefulness of the modified SLTHEN code for thermal hydraulic analysis and design of HYPER under the conceptual design stage. In order to assess intra-assembly heat transfer, subchannel analyses were implemented for two types of 7 assemblies; 1) artificial 7 fuel assemblies to maximize intra-assembly heat transfer, 2) central 7 fuel assemblies in the HYPER reference core. The results showed that the modified SLTHEN can reasonably simulate intra-heat transfer and the amount of intra-assembly heat transfer is not so large in HYPER conditions. Particularly, intra-heat transfer did not affect the maximum coolant and the maximum cladding temperatures which are major parameters in conceptual core designs. The capability of full core thermal hydraulic analysis was confirmed by the analysis of 45 fuel assemblies in 1/6 HYPER core at the first cycle. The SLTHEN predicted that the reference design parameters are acceptable in terms of thermal

SOURCE OF BURNUP VALUES FOR COMMERCIAL SPENT NUCLEAR FUEL ASSEMBLIES

International Nuclear Information System (INIS)

BSC

2004-01-01

, initial 235 U enrichment, and time of discharge from the reactor as well as the assigned burnup, but the distribution. of burnup axially along the assembly length is not provided. The axial burnup profile is maintained within acceptable bounds by the operating conditions of the nuclear reactor and is calculated during preparations to reload a reactor, but the actual burnup profile is not measured. The axial burnup profile is important to the determination of the reactivity of a waste package, so a conservative evaluation of the calculated axial profiles for a large database of SNF has been performed. The product of the axial profile evaluation is a profile that is conservative. Thus, there is no need for physical measurement of the axial profile. The assembly identifier is legible on each SNF assembly and the utility records provide the associated characteristics of the assembly. The conservative methodologies used to determine the criticality loading curve for a waste package provide sufficient margin so that criticality safety is assured for preclosure operations even in the event of a misload. Consideration of misload effects for postclosure time periods is provided by the criticality Features, Events, and Processes (FEPs) analysis. The conservative approaches used to develop and apply the criticality loading curve are thus sufficiently robust that the utility assigned burnup is an adequate source of burnup values, and additional means of verification of assigned burnup through physical measurements are not needed

Analysis of irradiation temperature in fuel rods of OGL-1 fuel assembly

International Nuclear Information System (INIS)

Fukuda, Kousaku; Kobayashi, Fumiaki; Minato, Kazuo; Ikawa, Katsuichi; Iwamoto, Kazumi

1984-10-01

Irradiation temperature in the fuel rods of 5th OGL-1 fuel assembly was analysed by the system composed by STPDSP2 and TRUMP codes. As the measured input-data, following parameters were allowed for; circumferential heating distribution around the fuel rod, which was measured in the JMTR critical assembly, axial heating distribution through the fuel rod, ratio of peak heatings of three fuel rods, and pre- and post-irradiation outer radii of the fuel compacts and inner radii of the graphite sleeves, which had been measured in PIE of the 5th OGL-1 fuel assembly. In computation the axial distributions of helium coolant temperature through the fuel rod and the heating value of each fuel rod were, firstly, calculated as input data for TRUMP. The TRUMP calculation yielded the temperatures which were fitted in those measured by all of the thermo-couples installed in the fuel rods, by adjusting only the value of the surface heat transfer coefficient, and consequently, the temperatures in all portions of the fuel rod were obtained. The apparent heat transfer coefficient changed to 60% of the initial values in the middle period of irradiation. For this reduction it was deduced that shoot had covered the surface of the fuel rod during irradiation, which was confirmed in PIE. Beside it, several things were found in this analysis. (author)

Neutron Activation and Thermoluminescent Detector Responses to a Bare Pulse of the CEA Valduc SILENE Critical Assembly

Energy Technology Data Exchange (ETDEWEB)

Miller, Thomas Martin [ORNL; Isbell, Kimberly McMahan [ORNL; Lee, Yi-kang [French Atomic Energy Commission (CEA), Centre de Saclay, Gif sur Yvette; Gagnier, Emmanuel [French Atomic Energy Commission (CEA), Centre de Saclay, Gif sur Yvette; Authier, Nicolas [French Atomic Energy Commission (CEA), Centre de Valduc, Is-sur-Tille; Piot, Jerome [French Atomic Energy Commission (CEA), Centre de Valduc, Is-sur-Tille; Jacquet, Xavier [French Atomic Energy Commission (CEA), Centre de Valduc, Is-sur-Tille; Rousseau, Guillaume [French Atomic Energy Commission (CEA), Centre de Valduc, Is-sur-Tille; Reynolds, Kevin H. [Y-12 National Security Complex

2016-09-01

This benchmark experiment was conducted as a joint venture between the US Department of Energy (DOE) and the French Commissariat à l'Energie Atomique (CEA). Staff at the Oak Ridge National Laboratory (ORNL) in the US and the Centre de Valduc in France planned this experiment. The experiment was conducted on October 11, 2010 in the SILENE critical assembly facility at Valduc. Several other organizations contributed to this experiment and the subsequent evaluation, including CEA Saclay, Lawrence Livermore National Laboratory (LLNL), the Y-12 National Security Complex (NSC), Babcock International Group in the United Kingdom, and Los Alamos National Laboratory (LANL). The goal of this experiment was to measure neutron activation and thermoluminescent dosimeter (TLD) doses from a source similar to a fissile solution critical excursion. The resulting benchmark can be used for validation of computer codes and nuclear data libraries as required when performing analysis of criticality accident alarm systems (CAASs). A secondary goal of this experiment was to qualitatively test performance of two CAAS detectors similar to those currently and formerly in use in some US DOE facilities. The detectors tested were the CIDAS MkX and the Rocky Flats NCD-91. These detectors were being evaluated to determine whether they would alarm, so they were not expected to generate benchmark quality data.

Supplementary neutron flux calculations for the ORNL pool critical assembly pressure vessel facility

Energy Technology Data Exchange (ETDEWEB)

Maerker, R.E.; Maudlin, P.J.

1981-02-01

A three-dimensional Monte Carlo calculation was performed to estimate the neutron flux in the 8/7 configuration of the ORNL Pool Critical Assembly Pressure Vessel Facility. The calculational tool was the multigroup transport code MORSE operated in the adjoint mode. The MORSE flux results compared well with those using a previously adopted procedure for constructing a three-dimensional flux from one- and two-dimensional discrete ordinates calculations using the DOT-IV code. This study concluded that use of these discrete ordinates constructions in previous calculations is sufficiently accurate and does not account for the existing discrepancies between calculation and experiment.

Supplementary neutron flux calculations for the ORNL pool critical assembly pressure vessel facility

International Nuclear Information System (INIS)

Maerker, R.E.; Maudlin, P.J.

1981-02-01

A three-dimensional Monte Carlo calculation was performed to estimate the neutron flux in the 8/7 configuration of the ORNL Pool Critical Assembly Pressure Vessel Facility. The calculational tool was the multigroup transport code MORSE operated in the adjoint mode. The MORSE flux results compared well with those using a previously adopted procedure for constructing a three-dimensional flux from one- and two-dimensional discrete ordinates calculations using the DOT-IV code. This study concluded that use of these discrete ordinates constructions in previous calculations is sufficiently accurate and does not account for the existing discrepancies between calculation and experiment

Modification of the ANC Nodal Code for analysis of PWR assembly bow

International Nuclear Information System (INIS)

Franceschini, Fausto; Fetterman, Robert J.; Little, David C.

2008-01-01

Refueling operations at certain PWR cores have revealed fuel assemblies with assembly bow that was higher than expected. As the fuel assemblies bow, the gaps between assemblies change from the uniform nominal configuration. This causes a change in the water volume which affects neutron moderation and thereby power distribution, fuel depletion history, rod internal pressure, etc., with non-trivial impacts on the safety analysis. Westinghouse has developed a new methodology for incorporation of assembly bow in its reload safety analysis package. As part of the new process, the standard Westinghouse reactor physics tool for core analysis, the Advanced Nodal Code ANC, has been modified. The modified ANC, ANCGAP, enables explicit treatment of three-dimensional gap distributions in its neutronic calculations; its accuracy is similar to that of the standard ANC, as demonstrated through an extensive benchmark campaign conducted over a variety of fuel compositions and challenging gap configurations. These features make ANCGAP a crucial tool in the Westinghouse assembly bow package. (authors)

Modification of the ANC Nodal Code for analysis of PWR assembly bow

Energy Technology Data Exchange (ETDEWEB)

Franceschini, Fausto; Fetterman, Robert J.; Little, David C. [Westinghouse Electric Company LLC, Pittsburgh PA (United States)

2008-07-01

Refueling operations at certain PWR cores have revealed fuel assemblies with assembly bow that was higher than expected. As the fuel assemblies bow, the gaps between assemblies change from the uniform nominal configuration. This causes a change in the water volume which affects neutron moderation and thereby power distribution, fuel depletion history, rod internal pressure, etc., with non-trivial impacts on the safety analysis. Westinghouse has developed a new methodology for incorporation of assembly bow in its reload safety analysis package. As part of the new process, the standard Westinghouse reactor physics tool for core analysis, the Advanced Nodal Code ANC, has been modified. The modified ANC, ANCGAP, enables explicit treatment of three-dimensional gap distributions in its neutronic calculations; its accuracy is similar to that of the standard ANC, as demonstrated through an extensive benchmark campaign conducted over a variety of fuel compositions and challenging gap configurations. These features make ANCGAP a crucial tool in the Westinghouse assembly bow package. (authors)

Feasibility study on heterogeneous method in criticality calculations

International Nuclear Information System (INIS)

Prati, A.

1977-01-01

The criticality of finite heterogeneous assemblies is analysed by the heterogeneous methods employing the Eigen-function analysis. The moderation is treated by the Fermi age theory. The system is analysed in two dimensional rectangular coordinates. The criticality and the fluxes are determined for systems with small and large number of fuel rods. The convergence and the residual error in the modal analysis are discussed. (author)

Assembly-level analysis of heterogeneous Th–Pu PWR fuel

International Nuclear Information System (INIS)

Zainuddin, Nurjuanis Zara; Parks, Geoffrey T.; Shwageraus, Eugene

2017-01-01

Highlights: • We directly compare homogeneous and heterogeneous Th–Pu fuel. • Examine whether there is an increase in Pu incineration in the latter. • Homogeneous fuel was able to achieve much higher Pu incineration. • In the heterogeneous case, U-233 breeding is faster (larger power fraction), thus decreasing incineration of Pu. - Abstract: This study compares homogeneous and heterogeneous thorium–plutonium (Th–Pu) fuel assemblies (with high Pu content – 20 wt%), and examines whether there is an increase in Pu incineration in the latter. A seed-blanket configuration based on the Radkowsky thorium reactor concept is used for the heterogeneous assembly. This separates the thorium blanket from the uranium seed, or in this case a plutonium seed. The seed supplies neutrons to the subcritical thorium blanket which encourages the in situ breeding and burning of "2"3"3U, allowing the fuel to stay critical for longer, extending burnup of the fuel. While past work on Th–Pu seed-blanket units shows superior Pu incineration compared to conventional U–Pu mixed oxide fuel, there is no literature to date that directly compares the performance of homogeneous and heterogeneous Th–Pu assembly configurations. Use of exactly the same fuel loading for both configurations allows the effects of spatial separation to be fully understood. It was found that the homogeneous fuel with and without burnable poisons was able to achieve much higher Pu incinerations than the heterogeneous fuel configurations, while still attaining a reasonably high discharge burnup. This is because in the heterogeneous cases, "2"3"3U breeding is faster, thereby contributing to a much larger fraction of total power produced by the assembly. In contrast, "2"3"3U build-up is slower in the homogeneous case and therefore Pu burning is greater. This "2"3"3U begins to contribute a significant fraction of power produced only towards the end of life, thus extending criticality, allowing more Pu to

Fast critical experiments in FCA and their analysis

International Nuclear Information System (INIS)

Hirota, Jitsuya

1984-02-01

JAERI Fast Critical Facility FCA went critical for the first time in April, 1967. Since then, critical experiments and their analysis were carried out on thirty-five assemblies until march, 1982. This report summarizes many achievements obtained in these fifteen years and points out disagreements observed between the calculation and experiment for further studies. A series of mock-up experiments for Experimental Fast Reactor JOYO, a theoretical and numerical study of adjustment of group constants by using integral data and a development of proton-recoil counter system for fast neutron spectrum measurement won high praise. Studies of Doppler effect of structural materials, effect of fission product accumulation on sodium-void worth, axially heterogeneous core and actinide cross sections attracted world-side attention. Significant contributions were also made to Prototype Fast Breeder Reactor MONJU through the partial mock-up experiments. Disagreements between the calculation and experiment were observed in the following items; reaction rate distribution and reactivity worth of B 4 C absorber in radial blanket, central reactivity worth in core with reflector, plate/pin fuel heterogeneity effect on criticality, sodium-void effect in central core region, Doppler effect of structural materials, core neutron spectrum near large resonances of iron and oxygen, effect of fission product accumulation on sodium-void worth, physics property of heterogeneous core, reactivity change resulted from fuel slumping and so on. Further efforts should be made to solve these disagreements through recalculating the experimental results with newly developed data and methods and carrying out the experiments intended to identify the cause of disagreement. (author)

Criticality safety study of dry spent fuel cask loaded with increased enrichment fuel

International Nuclear Information System (INIS)

Bznuni, S.; Baghdasaryan, N.; Amirjanyan, A.

2013-01-01

Existing Dry Spent Fuel Casks (DSC) for transporting and storing of Armenian NPP fuel was licensed for WWER-440 fuel assemblies with 3.6% enrichment. Having in mind that ANPP introduced new fuel assemblies with increased enrichment (3.82 %) re-assessment of criticality safety analysis for DSC is required. Criticality safety analysis of DSC was performed by KENO-VI program using 238-GROUP ENDF/B-VII.0 LIBRARY (V7-238). Results of analysis showed that additional 8 borated racks for fuel assemblies should be included in the design of DSC. In addition feasibility study was performed to find out level of burnup-credit approach implementation to keep current design of DSC unchanged. Burnup-credit analysis was performed by STARBUCS program using axial burnup profiles from Armenian NPP neutronics analysis carried out by BIPR code. (authors)

Applying critical analysis - main methods

Directory of Open Access Journals (Sweden)

Miguel Araujo Alonso

2012-02-01

Full Text Available What is the usefulness of critical appraisal of literature? Critical analysis is a fundamental condition for the correct interpretation of any study that is subject to review. In epidemiology, in order to learn how to read a publication, we must be able to analyze it critically. Critical analysis allows us to check whether a study fulfills certain previously established methodological inclusion and exclusion criteria. This is frequently used in conducting systematic reviews although eligibility criteria are generally limited to the study design. Critical analysis of literature and be done implicitly while reading an article, as in reading for personal interest, or can be conducted in a structured manner, using explicit and previously established criteria. The latter is done when formally reviewing a topic.

Launch and Assembly Reliability Analysis for Human Space Exploration Missions

Science.gov (United States)

Cates, Grant; Gelito, Justin; Stromgren, Chel; Cirillo, William; Goodliff, Kandyce

2012-01-01

NASA's future human space exploration strategy includes single and multi-launch missions to various destinations including cis-lunar space, near Earth objects such as asteroids, and ultimately Mars. Each campaign is being defined by Design Reference Missions (DRMs). Many of these missions are complex, requiring multiple launches and assembly of vehicles in orbit. Certain missions also have constrained departure windows to the destination. These factors raise concerns regarding the reliability of launching and assembling all required elements in time to support planned departure. This paper describes an integrated methodology for analyzing launch and assembly reliability in any single DRM or set of DRMs starting with flight hardware manufacturing and ending with final departure to the destination. A discrete event simulation is built for each DRM that includes the pertinent risk factors including, but not limited to: manufacturing completion; ground transportation; ground processing; launch countdown; ascent; rendezvous and docking, assembly, and orbital operations leading up to trans-destination-injection. Each reliability factor can be selectively activated or deactivated so that the most critical risk factors can be identified. This enables NASA to prioritize mitigation actions so as to improve mission success.

Formation of clusters composed of C60 molecules via self-assembly in critical fluids

International Nuclear Information System (INIS)

Fukuda, Takahiro; Ishii, Koji; Kurosu, Shunji; Whitby, Raymond; Maekawa, Toru

2007-01-01

Fullerenes are promising candidates for intelligent, functional nanomaterials because of their unique mechanical, electronic and chemical properties. However, it is necessary to invent some efficient but relatively simple methods of producing structures composed of fullerenes for the development of nanomechatronic, nanoelectronic and biochemical devices and sensors. In this paper, we show that various structures such as straight fibres, networks formed by fibres, wide sheets and helical structures, which are composed of C 60 molecules, are created by placing C 60 -crystals in critical ethane, carbon dioxide and xenon even though C 60 molecules do not dissolve or disperse in the above fluids. It is supposed, judging by the intermolecular potentials between C 60 and C 60 , between C 60 and ethane, and between ethane and ethane, that C 60 -clusters grow with the assistance of solvent molecules, which are trapped between C 60 molecules under critical conditions. This room-temperature self-assembly cluster growth process in critical fluids may open up a new methodology of forming structures built up with fullerenes without the need for any ultra-fine processing technologies

Comparative Neutronics Analysis of DIMPLE S06 Criticality Benchmark with Contemporary Reactor Core Analysis Computer Code Systems

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Wonkyeong Kim

2015-01-01

Full Text Available A high-leakage core has been known to be a challenging problem not only for a two-step homogenization approach but also for a direct heterogeneous approach. In this paper the DIMPLE S06 core, which is a small high-leakage core, has been analyzed by a direct heterogeneous modeling approach and by a two-step homogenization modeling approach, using contemporary code systems developed for reactor core analysis. The focus of this work is a comprehensive comparative analysis of the conventional approaches and codes with a small core design, DIMPLE S06 critical experiment. The calculation procedure for the two approaches is explicitly presented in this paper. Comprehensive comparative analysis is performed by neutronics parameters: multiplication factor and assembly power distribution. Comparison of two-group homogenized cross sections from each lattice physics codes shows that the generated transport cross section has significant difference according to the transport approximation to treat anisotropic scattering effect. The necessity of the ADF to correct the discontinuity at the assembly interfaces is clearly presented by the flux distributions and the result of two-step approach. Finally, the two approaches show consistent results for all codes, while the comparison with the reference generated by MCNP shows significant error except for another Monte Carlo code, SERPENT2.

Critical experiments simulating accidental water immersion of highly enriched uranium dioxide fuel elements

International Nuclear Information System (INIS)

Ponomarev-Stepnoi, N.N.; Glushkov, L.S.

2003-01-01

The paper focuses on experimental analysis of nuclear criticality safety at accidental water immersion of fuel elements of the Russian TOPAZ-2 space nuclear power system reactor. The structure of water-moderated heterogeneous critical assemblies at the NARCISS facility is described in detail, including sizes, compositions, densities of materials of the main assembly components for various core configurations. Critical parameters of the assemblies measured for varying number of fuel elements, height of fuel material in fuel elements and their arrangement in the water moderator with a uniform or variable spacing are presented. It has been found from the experiments that at accidental water immersion of fuel elements involved, the minimum critical mass equal to approximately 20 kg of uranium dioxide is achieved at 31-37 fuel elements. The paper gives an example of a physical model of the water-moderated heterogeneous critical assembly with a detailed characterization of its main components that can be used for calculations using different neutronic codes, including Monte Carlo ones. (author)

Onset of self-assembly

International Nuclear Information System (INIS)

Chitanvis, S.M.

1998-01-01

We have formulated a theory of self-assembly based on the notion of local gauge invariance at the mesoscale. Local gauge invariance at the mesoscale generates the required long-range entropic forces responsible for self-assembly in binary systems. Our theory was applied to study the onset of mesostructure formation above a critical temperature in estane, a diblock copolymer. We used diagrammatic methods to transcend the Gaussian approximation and obtain a correlation length ξ∼(c-c * ) -γ , where c * is the minimum concentration below which self-assembly is impossible, c is the current concentration, and γ was found numerically to be fairly close to 2/3. The renormalized diffusion constant vanishes as the critical concentration is approached, indicating the occurrence of critical slowing down, while the correlation function remains finite at the transition point. copyright 1998 The American Physical Society

Extrapolated experimental critical parameters of unreflected and steel-reflected massive enriched uranium metal spherical and hemispherical assemblies

International Nuclear Information System (INIS)

Rothe, R.E.

1997-12-01

Sixty-nine critical configurations of up to 186 kg of uranium are reported from very early experiments (1960s) performed at the Rocky Flats Critical Mass Laboratory near Denver, Colorado. Enriched (93%) uranium metal spherical and hemispherical configurations were studied. All were thick-walled shells except for two solid hemispheres. Experiments were essentially unreflected; or they included central and/or external regions of mild steel. No liquids were involved. Critical parameters are derived from extrapolations beyond subcritical data. Extrapolations, rather than more precise interpolations between slightly supercritical and slightly subcritical configurations, were necessary because experiments involved manually assembled configurations. Many extrapolations were quite long; but the general lack of curvature in the subcritical region lends credibility to their validity. In addition to delayed critical parameters, a procedure is offered which might permit the determination of prompt critical parameters as well for the same cases. This conjectured procedure is not based on any strong physical arguments

Thermohydraulic analysis of assemblies containing up to 2/7 fuel rods

International Nuclear Information System (INIS)

Ferreira, W.J.; Luz, M.

1985-01-01

The COBRA IV-I computer code was tested using data from the Fast Flux Test Facility. Then this code was applied to the analysis of fuel assemblies from the Binary Breeder Reactor. Previously this analysis was carried out using the COBRA III-C code which allows only for the calculations of fuel assemblies having seven fuel pins. The COBRA IV-I permits the calculation of fuel assemblies containing up to 217 fuel pins and the inclusion of blanket and shielding effects. (F.E.) [pt

Analysis of benchmark critical experiments with ENDF/B-VI data sets

International Nuclear Information System (INIS)

Hardy, J. Jr.; Kahler, A.C.

1991-01-01

Several clean critical experiments were analyzed with ENDF/B-VI data to assess the adequacy of the data for U 235 , U 238 and oxygen. These experiments were (1) a set of homogeneous U 235 -H 2 O assemblies spanning a wide range of hydrogen/uranium ratio, and (2) TRX-1, a simple, H 2 O-moderated Bettis lattice of slightly-enriched uranium metal rods. The analyses used the Monte Carlo program RCP01, with explicit three-dimensional geometry and detailed representation of cross sections. For the homogeneous criticals, calculated k crit values for large, thermal assemblies show good agreement with experiment. This supports the evaluated thermal criticality parameters for U 235 . However, for assemblies with smaller H/U ratios, k crit values increase significantly with increasing leakage and flux-spectrum hardness. These trends suggest that leakage is underpredicted and that the resonance eta of the ENDF/B-VI U 235 is too large. For TRX-1, reasonably good agreement is found with measured lattice parameters (reaction-rate ratios). Of primary interest is rho28, the ratio of above-thermal to thermal U 238 capture. Calculated rho28 is 2.3 (± 1.7) % above measurement, suggesting that U 238 resonance capture remains slightly overpredicted with ENDF/B-VI. However, agreement is better than observed with earlier versions of ENDF/B

Test calculations of physical parameters of the TRX,BETTIS and MIT critical assemblies according to the TRIFON program

International Nuclear Information System (INIS)

Kochurov, B.P.

1980-01-01

Results of calculations of physical parameters characterizing the TRX, MIT and BETTIS critical assemblies obtained according to the program TRIFON are presented. The program TRIFON permits to calculate the space-energy neutron distribution in the multigroup approximation in a multizone cylindrical cell. Results of comparison of the TRX, BETTIS and MIT crytical assembly parameters with experimental data and calculational results according to the Monte Carlo method are presented as well. Deviations of the parameters are in the range of 1.5-2 of experimental errors. Data on the interference of uranium 238 levels in the resonant neutron absorption in the cell are given [ru

Three-Dimensional Assembly Tolerance Analysis Based on the Jacobian-Torsor Statistical Model

Directory of Open Access Journals (Sweden)

Peng Heping

2017-01-01

Full Text Available The unified Jacobian-Torsor model has been developed for deterministic (worst case tolerance analysis. This paper presents a comprehensive model for performing statistical tolerance analysis by integrating the unified Jacobian-Torsor model and Monte Carlo simulation. In this model, an assembly is sub-divided into surfaces, the Small Displacements Torsor (SDT parameters are used to express the relative position between any two surfaces of the assembly. Then, 3D dimension-chain can be created by using a surface graph of the assembly and the unified Jacobian-Torsor model is developed based on the effect of each functional element on the whole functional requirements of products. Finally, Monte Carlo simulation is implemented for the statistical tolerance analysis. A numerical example is given to demonstrate the capability of the proposed method in handling three-dimensional assembly tolerance analysis.

Core fuel management using TVS-2M fuel assembly and economic analysis

International Nuclear Information System (INIS)

Xu Min; Wang Hongxia; Li Youyi

2014-01-01

To improve the economic efficiency, TVS-2M fuel assembly was considered to apply in Tianwan Nuclear Power Plant units 3, 4. Using KASKAD program package, a preliminary research and design was carried out for the Tianwan Nuclear Power Plant loading TVS-2M fuel assembly from the first cycle to equilibrium cycle. An improved fuel management program was obtained, and the economic analysis of the two fuel management programs with or without TVS-2M assembly was studied. The analysis results show that TVS-2M fuel assembly can improve the economic efficiency of the plant remarkably. (authors)

Network spandrels reflect ecological assembly.

Science.gov (United States)

Maynard, Daniel S; Serván, Carlos A; Allesina, Stefano

2018-03-01

Ecological networks that exhibit stable dynamics should theoretically persist longer than those that fluctuate wildly. Thus, network structures which are over-represented in natural systems are often hypothesised to be either a cause or consequence of ecological stability. Rarely considered, however, is that these network structures can also be by-products of the processes that determine how new species attempt to join the community. Using a simulation approach in tandem with key results from random matrix theory, we illustrate how historical assembly mechanisms alter the structure of ecological networks. We demonstrate that different community assembly scenarios can lead to the emergence of structures that are often interpreted as evidence of 'selection for stability'. However, by controlling for the underlying selection pressures, we show that these assembly artefacts-or spandrels-are completely unrelated to stability or selection, and are instead by-products of how new species are introduced into the system. We propose that these network-assembly spandrels are critically overlooked aspects of network theory and stability analysis, and we illustrate how a failure to adequately account for historical assembly can lead to incorrect inference about the causes and consequences of ecological stability. © 2018 John Wiley & Sons Ltd/CNRS.

Reactor laboratory course for Korean under-graduate students in Kyoto University Critical Assembly (KUGSiKUCA)

International Nuclear Information System (INIS)

Pyeon, Cheol Ho; Misawa, Tsuyoshi; Unesaki, Hironobu; Ichihara, Chihiro; Shiroya, Seiji; Whang, Joo Ho; Kim, Myung Hyun

2005-01-01

The Reactor Laboratory Course for Korean Under-Graduate Students has been carried out at Kyoto University Critical Assembly of Japan. This course has been launched from fiscal year 2003 and has been founded by Ministry of Science and Technology of Korean Government. Since then, the total number of 43 Korean under-graduate students, who have majored in nuclear engineering of 6 universities in all over the Korea, has been taken part in this course. The reactor physics experiments have been performed in this course, such as Approach to criticality, Control rod calibration, Measurement of neutron flux and power calibration, and Educational reactor operation. As technical tour of Japan, nuclear site tour has been taken during their stay in Japan, such as PWR, FBR, nuclear fuel company and some institutes

Transmutation of uranium and thorium in the particle field of the Quinta sub-critical assembly

Science.gov (United States)

Hashemi-Nezhad, S. R.; Asquith, N. L.; Voronko, V. A.; Sotnikov, V. V.; Zhadan, Alina; Zhuk, I. V.; Potapenko, A.; Husak, Krystsina; Chilap, V.; Adam, J.; Baldin, A.; Berlev, A.; Furman, W.; Kadykov, M.; Khushvaktov, J.; Kudashkin, I.; Mar'in, I.; Paraipan, M.; Pronskih, V.; Solnyshkin, A.; Tyutyunnikov, S.

2018-03-01

The fission rates of natural uranium and thorium were measured in the particle field of Quinta, a 512 kg natural uranium target-blanket sub-critical assembly. The Quinta assembly was irradiated with deuterons of energy 4 GeV from the Nuclotron accelerator of the Joint Institute for Nuclear Research (JINR), Dubna, Russia. Fission rates of uranium and thorium were measured using Gamma spectroscopy and fission track techniques. The production rate of 239Np was also measured. The obtained experimental results were compared with Monte Carlo predictions using the MCNPX 2.7 code employing the physics and fission-evaporation models of INCL4-ABLA, CEM03.03 and LAQGSM03.03. Some of the neutronic characteristics of the Quinta are compared with the "Energy plus Transmutation (EpT)" subcritical assembly, which is composed of a lead target and natU blanket. This comparison clearly demonstrates the importance of target material, neutron moderator and reflector types on the performance of a spallation neutron driven subcritical system. As the dimensions of the Quinta are very close to those of an optimal multi-rod-uranium target, the experimental and Monte Carlo calculation results presented in this paper provide insights on the particle field within a uranium target as well as in Accelerator Driven Systems in general.

Multi-scale coarse-graining for the study of assembly pathways in DNA-brick self-assembly

Science.gov (United States)

Fonseca, Pedro; Romano, Flavio; Schreck, John S.; Ouldridge, Thomas E.; Doye, Jonathan P. K.; Louis, Ard A.

2018-04-01

Inspired by recent successes using single-stranded DNA tiles to produce complex structures, we develop a two-step coarse-graining approach that uses detailed thermodynamic calculations with oxDNA, a nucleotide-based model of DNA, to parametrize a coarser kinetic model that can reach the time and length scales needed to study the assembly mechanisms of these structures. We test the model by performing a detailed study of the assembly pathways for a two-dimensional target structure made up of 334 unique strands each of which are 42 nucleotides long. Without adjustable parameters, the model reproduces a critical temperature for the formation of the assembly that is close to the temperature at which assembly first occurs in experiments. Furthermore, the model allows us to investigate in detail the nucleation barriers and the distribution of critical nucleus shapes for the assembly of a single target structure. The assembly intermediates are compact and highly connected (although not maximally so), and classical nucleation theory provides a good fit to the height and shape of the nucleation barrier at temperatures close to where assembly first occurs.

Single-Molecule Analysis for RISC Assembly and Target Cleavage.

Science.gov (United States)

Sasaki, Hiroshi M; Tadakuma, Hisashi; Tomari, Yukihide

2018-01-01

RNA-induced silencing complex (RISC) is a small RNA-protein complex that mediates silencing of complementary target RNAs. Biochemistry has been successfully used to characterize the molecular mechanism of RISC assembly and function for nearly two decades. However, further dissection of intermediate states during the reactions has been warranted to fill in the gaps in our understanding of RNA silencing mechanisms. Single-molecule analysis with total internal reflection fluorescence (TIRF) microscopy is a powerful imaging-based approach to interrogate complex formation and dynamics at the individual molecule level with high sensitivity. Combining this technique with our recently established in vitro reconstitution system of fly Ago2-RISC, we have developed a single-molecule observation system for RISC assembly. In this chapter, we summarize the detailed protocol for single-molecule analysis of chaperone-assisted assembly of fly Ago2-RISC as well as its target cleavage reaction.

Safe operation of research reactors and critical assemblies. Code of practice and annexes. 1984 ed

International Nuclear Information System (INIS)

1984-01-01

The safe operation of research reactors and critical assemblies (hereafter termed 'reactors') requires proper design, construction, management and supervision. This Code of Practice deals mainly with management and supervision. The provisions of the Code apply to the whole life of the reactor, including modification, updating and upgrading. The Code may be subject to revision in the light of experience and the state of technology. The Code is aimed at defining minimum requirements for the safe operation of reactors. Emphasis is placed on which safety requirements should be met rather than on specifying how these requirements may be met. The Code also provides guidance and information to persons and authorities responsible for the operation of reactors. The Code recommends that documents dealing with the operation of reactors and including safety analyses be prepared and submitted for review and approval to a regulatory body. Operation would be authorized on the understanding that it would comply with limits and conditions designed to ensure safety. The Code covers a wide range of reactor types, which gives rise to a variety of safety issues. Safety issues applicable to specific reactor types only (e.g. fast reactors) are not necessarily covered in this Code. Some of the recommendations in the Code are not directly applicable to critical assemblies. A recommendation may therefore be interpreted according to the type of reactor concerned. In such cases the words 'adequate' and 'appropriate' are used to mean 'adequate' or 'appropriate' for the type of reactor under consideration.

Rapid method of calculating the fluence and spectrum of neutrons from a critical assembly and the resulting dose

International Nuclear Information System (INIS)

Bessis, J.

1977-01-01

The proposed calculation method is unsophisticated but rapid. The first part (computer code CRITIC), which is based on the Fermi age equation, evaluates the number of neutrons per fission emitted from a moderated critical assembly and their energy spectrum. The second part (computer code NARCISSE), which uses the current albedo for concrete, evaluates the product of neutron reflection on the walls and calculates the fluence resulting at any point in the room and its energy distribution by 21 groups. The results obtained are shown to compare satisfactorily with those obtained through the use of a Monte Carlo program

Three-Dimensional Heat Transfer Analysis of Metal Fasteners in Roofing Assemblies

Directory of Open Access Journals (Sweden)

Manan Singh

2016-11-01

Full Text Available Heat transfer analysis was performed on typical roofing assemblies using HEAT3, a three-dimensional heat transfer analysis software. The difference in heat transferred through the roofing assemblies considered is compared between two cases—without any steel fasteners and with steel fasteners. In the latter case, the metal roofing fasteners were arranged as per Factor Mutual Global (FMG approvals, in the field, perimeter, and corner zones of the roof. The temperature conditions used for the analysis represented summer and winter conditions for three separate Climate Zones (CZ namely Climate Zone 2 or CZ2 represented by Orlando, FL; CZ3 represented by Atlanta, GA; and CZ6 zone represented by St. Paul, MN. In all the climatic conditions, higher energy transfer was observed with increase in the number of metal fasteners attributed to high thermal conductivity of metals as compared to the insulation and other materials used in the roofing assembly. This difference in heat loss was also quantified in the form of percentage change in the overall or effective insulation of the roofing assembly for better understanding of the practical aspects. Besides, a comparison of 2D heat transfer analysis (using THERM software and 3D analysis using HEAT3 is also discussed proving the relevance of 3D over 2D heat transfer analysis.

Analysis of the Retained Gas Sample (RGS) Extruder Assembly

International Nuclear Information System (INIS)

Coverdell, B.L.

1995-09-01

In order for the Retained Gas Sample (RGS) Extruder Assembly to be safely used it was determined by the cognizant engineer that analysis was necessary. The use of the finite-element analysis (FEA) progarm COSMOS/M version 1.71 permitted a quick, easy, and detailed stress analysis of the RGS Extruder Assembly. The FEA model is a three dimensional model using the SHELL4T element type. From the results of the FEA, the cognizant engineer determined that the RGS extruder would be rated at 10,000 lbf and load tested to 12,000 lbf. The respective input and output files for the model are EXTR02.GFM and EXTR02.OUT and can be found on the attached tape

DOE Lab-to-Lab MPC ampersand A workshop for cooperative tasks with Russian institutes: Focus on critical assemblies and item facilities

International Nuclear Information System (INIS)

Bieber, A.M. Jr.; Fishbone, L.G.; Kato, W.Y.; Lazareth, O.W.; Suda, S.C.; Garcia, D.; Haga, R.

1995-01-01

Seventeen Russian scientists and engineers representing five different institutes participated in a Workshop on material control and accounting as part of the US-Russian Lab-to-Lab Cooperative Program in Nuclear Materials Protection, Control, and Accounting (MPC ampersand A). In addition to presentations and discussions, the Workshop included an exercise at Brookhaven National Laboratory (BNL) and demonstrations at the Zero Power Physics Reactor (critical-assembly facility) of Argonne National Laboratory-West (ANL-W). The Workshop particularly emphasized procedures for physical inventory-taking at critical assemblies and item facilities, with associated supporting techniques and methods. By learning these topics and applying the methods and experience at their own institutes, the Russian scientists and engineers will be able to determine and verify nuclear material inventories based on sound procedures, including measurements. This will constitute a significant enhancement to MPC ampersand A at the Russian institutes

The Texts of the Instruments connected with the Agency's Assistance to Finland in Establishing a Sub-Critical Assemblies Project

International Nuclear Information System (INIS)

1964-01-01

The text of the Project Agreement between the Agency and the Government of Finland in connection with the Agency's assistance to that Government in establishing a sub-critical assemblies project is reproduced in Part I of this document for the information of all Members. This Agreement entered into force on 30 July 1963

The Texts of the Instruments connected with the Agency's Assistance to Finland in Establishing a Sub-Critical Assemblies Project

Energy Technology Data Exchange (ETDEWEB)

NONE

1964-02-10

The text of the Project Agreement between the Agency and the Government of Finland in connection with the Agency's assistance to that Government in establishing a sub-critical assemblies project is reproduced in Part I of this document for the information of all Members. This Agreement entered into force on 30 July 1963.

Safety analysis of the Los Alamos Critical Experiments Facility. Volume II

International Nuclear Information System (INIS)

Paxton, H.C.

1976-04-01

The Los Alamos critical assembly layout is designed to facilitate personnel protection by means of remote operation and stringent procedural controls during nonoperating periods. Public protection is straightforward because of the small fission-product inventory, essentially ambient pressures, and moderate temperatures

Detailed Structural Analysis of Critical Wendelstein 7-X Magnet System Components

International Nuclear Information System (INIS)

Egorov, K.

2006-01-01

The Wendelstein 7-X (W7-X) stellarator experiment is presently under construction and assembly in Greifswald, Germany. The goal of the experiment is to verify that the stellarator magnetic confinement concept is a viable option for a fusion reactor. The complex W7-X magnet system requires a multi-level approach to structural analysis for which two types of finite element models are used: Firstly, global models having reasonably coarse meshes with a number of simplifications and assumptions, and secondly, local models with detailed meshes of critical regions and elements. Widely known sub-modelling technique with boundary conditions extracted from the global models is one of the approaches for local analysis with high assessment efficiency. In particular, the winding pack (WP) of the magnet coils is simulated in the global model as a homogeneous orthotropic material with effective mechanical characteristic representing its real composite structure. This assumption allows assessing the whole magnet system in terms of general structural factors like forces and moments on the support elements, displacements of the main components, deformation and stress in the coil casings, etc. In a second step local models with a detailed description of more critical WP zones are considered in order to analyze their internal components like conductor jackets, turn insulation, etc. This paper provides an overview of local analyses of several critical W7-X magnet system components with particular attention on the coil winding packs. (author)

Analysis of assembly serial number usage in domestic light-water reactors

International Nuclear Information System (INIS)

Reich, W.J.; Moore, R.S.

1991-05-01

Domestic light-water reactor (LWR) fuel assemblies are identified by a serial number that is placed on each assembly. These serial numbers are used as identifiers throughout the life of the fuel. The uniqueness of assembly serial numbers is important in determining their effectiveness as unambiguous identifiers. The purpose of this study is to determine what serial numbering schemes are used, the effectiveness of these schemes, and to quantify how many duplicate serial numbers occur on domestic LWR fuel assemblies. The serial numbering scheme adopted by the American National Standards Institute (ANSI) ensures uniqueness of assembly serial numbers. The latest numbering scheme adopted by General Electric (GE), was also found to be unique. Analysis of 70,971 fuel assembly serial numbers from permanently discharged fuel identified 11,948 serial number duplicates. Three duplicate serial numbers were found when analysis focused on duplication within the individual fuel inventory at each reactor site, but these were traced back to data entry errors and will be corrected by the Energy Information Administration (EIA). There were also three instances where the serial numbers used to identify assemblies used for hot cell studies differed from the serial numbers reported to the EIA. It is recommended that fuel fabricators and utilities adhere to the ANSI serial numbering scheme to ensure serial number uniqueness. In addition, organizations collecting serial number information, should request that all known serial numbers physically attached or associated with each assembly be reported and identified by the corresponding number scheme. 10 refs., 5 tabs

Seismic analysis of fuel and target assemblies at a production reactor

International Nuclear Information System (INIS)

Braverman, J.I.; Wang, Y.K.

1991-01-01

This paper describes the unique modeling and analysis considerations used to assess the seismic adequacy of the fuel and target assemblies in a production reactor at Savannah River Site. This confirmatory analysis was necessary to provide assurance that the reactor can operate safely during a seismic event and be brought to a safe shutdown condition. The plant which was originally designed in the 1950's required to be assessed to more current seismic criteria. The design of the reactor internals and the magnitude of the structural responses enabled the use of a linear elastic dynamic analysis. A seismic analysis was performed using a finite element model consisting of the fuel and target assemblies, reactor tank, and a portion of the concrete structure supporting the reactor tank. The effects of submergence of the fuel and target assemblies in the water contained within the reactor tank can have a significant effect on their seismic response. Thus, the model included hydrodynamic fluid coupling effects between the assemblies and the reactor tank. Fluid coupling mass terms were based on formulations for solid bodies immersed in incompressible and frictionless fluids. The potential effects of gap conditions were also assessed in this evaluation. 5 refs., 6 figs., 1 tab

Development of CFD analysis method based on droplet tracking model for BWR fuel assemblies

International Nuclear Information System (INIS)

Onishi, Yoichi; Minato, Akihiko; Ichikawa, Ryoko; Mashara, Yasuhiro

2011-01-01

It is well known that the minimum critical power ratio (MCPR) of the boiling water reactor (BWR) fuel assembly depends on the spacer grid type. Recently, improvement of the critical power is being studied by using a spacer grid with mixing devices attaching various types of flow deflectors. In order to predict the critical power of the improved BWR fuel assembly, we have developed an analysis method based on the consideration of detailed thermal-hydraulic mechanism of annular mist flow regime in the subchannels for an arbitrary spacer type. The proposed method is based on a computational fluid dynamics (CFD) model with a droplet tracking model for analyzing the vapor-phase turbulent flow in which droplets are transported in the subchannels of the BWR fuel assembly. We adopted the general-purpose CFD software Advance/FrontFlow/red (AFFr) as the base code, which is a commercial software package created as a part of Japanese national project. AFFr employs a three-dimensional (3D) unstructured grid system for application to complex geometries. First, AFFr was applied to single-phase flows of gas in the present paper. The calculated results were compared with experiments using a round cellular spacer in one subchannel to investigate the influence of the choice of turbulence model. The analyses using the large eddy simulation (LES) and re-normalisation group (RNG) k-ε models were carried out. The results of both the LES and RNG k-ε models show that calculations of velocity distribution and velocity fluctuation distribution in the spacer downstream reproduce the experimental results qualitatively. However, the velocity distribution analyzed by the LES model is better than that by the RNG k-ε model. The velocity fluctuation near the fuel rod, which is important for droplet deposition to the rod, is also simulated well by the LES model. Then, to examine the effect of the spacer shape on the analytical result, the gas flow analyses with the RNG k-ε model were performed

Pendulum support of the W7-X plasma vessel: Design, tests, manufacturing, assembly, critical aspects, status

Energy Technology Data Exchange (ETDEWEB)

Missal, B., E-mail: bernd.missal@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Teilinstitut Greifswald, Wendelsteinstraße 1, D-17491 Greifswald (Germany); Leher, F.; Schiller, T. [MAN Diesel and Turbo SE, Werftstraße 17, 94469 Deggendorf (Germany); Friedrich, P. [Universität Rostock, FB Maschinenbau und Schiffstechnik, Albert-Einsteins-Straße 2, 18051 Rostock (Germany); Capriccioli, A. [ENEA Frascati, Fusion Technology Unit, Frascati (Italy)

2014-10-15

Highlights: • Plasma vessel support has to allow vertical adjustment and horizontal passive movement. • Planar sliding tables with PTFE do not fulfill all requirements. • Pendulums can fulfill all requirements. • Geometry and material of spherical bearings had to be optimized in calculations and tests. • Optimized pendulums were manufactured and assembled. - Abstract: The superconducting helical advanced stellarator Wendelstein 7-X (W7-X) is under construction at the Max-Planck-Institut für Plasmaphysik (IPP) in Greifswald, Germany. The three dimensional shape of plasma will be generated by 50 non-planar magnetic coils. The plasma vessel geometry follows exactly this three dimensional shape of plasma. To ensure the superconductivity of coils a cryo vacuum has to be generated. Therefore the coils and their support structure are enclosed within the outer vessel. Plasma vessel, coil structures and outer vessel have to be supported separately. This paper will describe the vertical supports of plasma vessel which have to fulfill two special requirements, vertical adjustability and horizontal mobility. These two tasks will be carried out by plasma vessel supports (PVS) with hydraulic cylinders, special sliding tables during assembly and pendulum supports during operating phase. The paper will give an overview of design, calculation, tests, fabrication, assembly, critical aspects and status of PVS.

Analysis of kyoto university reactor physics critical experiments using NCNSRC calculation methodology

International Nuclear Information System (INIS)

Amin, E.; Hathout, A.M.; Shouman, S.

1997-01-01

The kyoto university reactor physics experiments on the university critical assembly is used to benchmark validate the NCNSRC calculations methodology. This methodology has two lines, diffusion and Monte Carlo. The diffusion line includes the codes WIMSD4 for cell calculations and the two dimensional diffusion code DIXY2 for core calculations. The transport line uses the MULTIKENO-Code vax Version. Analysis is performed for the criticality, and the temperature coefficients of reactivity (TCR) for the light water moderated and reflected cores, of the different cores utilized in the experiments. The results of both Eigen value and TCR approximately reproduced the experimental and theoretical Kyoto results. However, some conclusions are drawn about the adequacy of the standard wimsd4 library. This paper is an extension of the NCNSRC efforts to assess and validate computer tools and methods for both Et-R R-1 and Et-MMpr-2 research reactors. 7 figs., 1 tab

Best estimate approach for the evaluation of critical heat flux phenomenon in the boiling water reactors

Energy Technology Data Exchange (ETDEWEB)

Kaliatka, Tadas; Kaliatka, Algirdas; Uspuras, Eudenijus; Vaisnoras, Mindaugas [Lithuanian Energy Institute, Kaunas (Lithuania); Mochizuki, Hiroyasu; Rooijen, W.F.G. van [Fukui Univ. (Japan). Research Inst. of Nuclear Engineering

2017-05-15

Because of the uncertainties associated with the definition of Critical Heat Flux (CHF), the best estimate approach should be used. In this paper the application of best-estimate approach for the analysis of CHF phenomenon in the boiling water reactors is presented. At first, the nodalization of RBMK-1500, BWR-5 and ABWR fuel assemblies were developed using RELAP5 code. Using developed models the CHF and Critical Heat Flux Ratio (CHFR) for different types of reactors were evaluated. The calculation results of CHF were compared with the well-known experimental data for light water reactors. The uncertainty and sensitivity analysis of ABWR 8 x 8 fuel assembly CHFR calculation result was performed using the GRS (Germany) methodology with the SUSA tool. Finally, the values of Minimum Critical Power Ratio (MCPR) were calculated for RBMK-1500, BWR-5 and ABWR fuel assemblies. The paper demonstrate how, using the results of sensitivity analysis, to receive the MCPR values, which covers all uncertainties and remains best estimated.

Evaluation of nine popular de novo assemblers in microbial genome assembly.

Science.gov (United States)

Forouzan, Esmaeil; Maleki, Masoumeh Sadat Mousavi; Karkhane, Ali Asghar; Yakhchali, Bagher

2017-12-01

Next generation sequencing (NGS) technologies are revolutionizing biology, with Illumina being the most popular NGS platform. Short read assembly is a critical part of most genome studies using NGS. Hence, in this study, the performance of nine well-known assemblers was evaluated in the assembly of seven different microbial genomes. Effect of different read coverage and k-mer parameters on the quality of the assembly were also evaluated on both simulated and actual read datasets. Our results show that the performance of assemblers on real and simulated datasets could be significantly different, mainly because of coverage bias. According to outputs on actual read datasets, for all studied read coverages (of 7×, 25× and 100×), SPAdes and IDBA-UD clearly outperformed other assemblers based on NGA50 and accuracy metrics. Velvet is the most conservative assembler with the lowest NGA50 and error rate. Copyright © 2017. Published by Elsevier B.V.

Critical experiment and analysis for nitride fuel fast reactor using FCA

International Nuclear Information System (INIS)

Andoh, Masaki; Iijima, Susumu; Okajima, Shigeaki; Sakurai, Takeshi; Oigawa, Hiroyuki

2000-03-01

As a research on FBR with new types of fuel, a series of experiments on a nitride fuel fast reactor was carried out at Fast Critical Assembly (FCA) to evaluate the calculation accuracy on the neutronic characteristics of the reactor. In this study, criticality, sample reactivity worth and sodium void reactivity worth were measured in the FCA XIX-2 core simulating a nitride fuel fast reactor and were analyzed using the standard analysis method for FCA fast reactor cores. The accuracy of the analysis on the effective multiplication factor was the same as those of the other FCA cores. For the plate sample reactivity worth, the calculation on the radial distribution of plutonium plate reactivity worth overestimated the measurement depending on the distance from the center of the core. For the sodium void reactivity worth, the calculation overestimated the experimental value 10 to 20% at the core center, while the overestimation was improved as the voided position was located at the core boundary. It was found that the transport effect was considerable even at the center of the core. It was considered that the calculation accuracy on the non-leakage term of the void reactivity worth and transport correction should be improved. (author)

Orientation-controlled parallel assembly at the air–water interface

International Nuclear Information System (INIS)

Park, Kwang Soon; Hoo, Ji Hao; Baskaran, Rajashree; Böhringer, Karl F

2012-01-01

This paper presents an experimental and theoretical study with statistical analysis of a high-yield, orientation-specific fluidic self-assembly process on a preprogrammed template. We demonstrate self-assembly of thin (less than few hundred microns in thickness) parts, which is vital for many applications in miniaturized platforms but problematic for today's pick-and-place robots. The assembly proceeds row-by-row as the substrate is pulled up through an air–water interface. Experiments and analysis are presented with an emphasis on the combined effect of controlled surface waves and magnetic force. For various gap values between a magnet and Ni-patterned parts, magnetic force distributions are generated using Monte Carlo simulation and employed to predict assembly yield. An analysis of these distributions shows that a gradual decline in yield following the probability density function can be expected with degrading conditions. The experimentally determined critical magnetic force is in good agreement with a derived value from a model of competing forces acting on a part. A general set of design guidelines is also presented from the developed model and experimental data. (paper)

The relevance of axial burn-up profiles for the criticality safety analysis of spent nuclear fuel in a final repository

International Nuclear Information System (INIS)

Kilger, R.; Gmal, B.; Moser, E.F.

2008-01-01

Due to inhomogeneous neutron flux and moderator density distributions in the reactor core, the burn-up of a nuclear fuel assembly is not homogeneous but shows an axial distribution, typically with lower partial burn-up and thus higher remaining reactivity at the fuel ends in particular at the assembly top end. Beyond a burn-up of about 15 to 20 GWd/tHM, the multiplication factor K of the whole assembly is dominated by this lower-burnt end regions, and is usually higher than for assuming a homogeneous uniform distribution of the averaged burn-up. This behaviour commonly referred to as positive ''end effect'' is well known in burn-up credit considerations for transportation and storage casks and is being investigated also in the context of criticality analyses for final disposition of spent nuclear fuel. Sign and value of the end effect depend on several parameters. Based on a generic model one may not conclude that criticality in a final repository is a likely or expected event, but nevertheless it draws the attention to the fact that criticality is not excluded per se but has to be considered in the analysis and probably has to be encountered by certain appropriate measures, maybe e.g. by limitation of the amount of fissile material inside one single cask, or a rigorous prove for prevention of water ingress. The authors also conclude that the higher partial reactivity of the fuel ends has to be accounted for carefully in more realistic analyses of post-closure scenarios with respect to criticality safety.

Criticality analysis for weapon disassembly at the Pantex Plant - part I: Bare pits

Energy Technology Data Exchange (ETDEWEB)

Knief, R.A. [Ogden Environmental & Energy Services, Albuquerque, NM (United States)

1997-06-01

This paper briefly describes criticality investigations for weapon assembly and dismantlement at the Pantex Plant. Results are summarized for calculations performed for safety analyses, radiological hazards assessments, and a study to justify the criticality alarm exemption. Pits and pits in containers were modeled in their most reactive configuration. Criticality calculations were performed with the KENO and MCNP code packages. Configurations involving bare pits were subcritical by a substantial amount even with very conservative model assumptions. Thus, it is concluded that a critical configuration involving the bare pits is not credible.

Nuclear Criticality Experimental Research Center (NCERC) Overview

Energy Technology Data Exchange (ETDEWEB)

Goda, Joetta Marie [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Grove, Travis Justin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hayes, David Kirk [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Myers, William L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Sanchez, Rene Gerardo [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-03-03

The mission of the National Criticality Experiments Research Center (NCERC) at the Device Assembly Facility (DAF) is to conduct experiments and training with critical assemblies and fissionable material at or near criticality in order to explore reactivity phenomena, and to operate the assemblies in the regions from subcritical through delayed critical. One critical assembly, Godiva-IV, is designed to operate above prompt critical. The Nuclear Criticality Experimental Research Center (NCERC) is our nation’s only general-purpose critical experiments facility and is only one of a few that remain operational throughout the world. This presentation discusses the history of NCERC, the general activities that makeup work at NCERC, and the various government programs and missions that NCERC supports. Recent activities at NCERC will be reviewed, with a focus on demonstrating how NCERC meets national security mission goals using engineering fundamentals. In particular, there will be a focus on engineering theory and design and applications of engineering fundamentals at NCERC. NCERC activities that relate to engineering education will also be examined.

Effects of existing evaluated nuclear data files on neutronics characteristics of the BFS-62-3A critical assembly benchmark model

International Nuclear Information System (INIS)

Semenov, Mikhail

2002-11-01

This report is continuation of studying of the experiments performed on BFS-62-3A critical assembly in Russia. The objective of work is definition of the cross section uncertainties on reactor neutronics parameters as applied to the hybrid core of the BN-600 reactor of Beloyarskaya NPP. Two-dimensional benchmark model of BFS-62-3A was created specially for these purposes and experimental values were reduced to it. Benchmark characteristics for this assembly are 1) criticality; 2) central fission rate ratios (spectral indices); and 3) fission rate distributions in stainless steel reflector. The effects of nuclear data libraries have been studied by comparing the results calculated using available modern data libraries - ENDF/B-V, ENDF/B-VI, ENDF/B-VI-PT, JENDL-3.2 and ABBN-93. All results were computed by Monte Carlo method with the continuous energy cross-sections. The checking of the cross sections of major isotopes on wide benchmark criticality collection was made. It was shown that ENDF/B-V data underestimate the criticality of fast reactor systems up to 2% Δk. As for the rest data, the difference between each other in criticality for BFS-62-3A is around 0.6% Δk. However, taking into account the results obtained for other fast reactor benchmarks (and steel-reflected also), it may conclude that the difference in criticality calculation results can achieve 1% Δk. This value is in a good agreement with cross section uncertainty evaluated for BN-600 hybrid core (±0.6% Δk). This work is related to the JNC-IPPE Collaboration on Experimental Investigation of Excess Weapons Grade Pu Disposition in BN-600 Reactor Using BFS-2 Facility. (author)

Seismic Response Analysis of Assembled Reactor Vessel Internals

International Nuclear Information System (INIS)

Je, Sang-Yun; Chang, Yoon-Suk; Kang, Sung-Sik

2015-01-01

RVIs (Reactor Vessel Internals) perform important safe-related functions such as upholding the nuclear fuel assembly as well as providing the coolant passage of the reactor core and supporting the control rod drive mechanism. Therefore, the components including RVIs have to be designed and constructed taking into account the structural integrity under various accident scenarios. The reliable seismic analysis is essentially demanded to maintain the safe-related functions of RVIs. In this study, a modal analysis was performed based on the previous researches to examine values of frequencies, mode shapes and participation factors. Subsequently, the structural integrity respecting to the earthquake was evaluated through a response spectrum analysis by using the output variables of modal analysis. In this study, the structural integrity of the assembled RVIs was carried out against the seismic event via the modal and response spectrum analyses. Even though 287MPa of the maximum stress value occurred at the connected region between UGS and CSA, which was lower than its allowable value. At present, fluid-structure interaction effects are being examined for further realistic simulation, which will be reported in the near future

Equilibrium polymerization models of re-entrant self-assembly

Science.gov (United States)

Dudowicz, Jacek; Douglas, Jack F.; Freed, Karl F.

2009-04-01

As is well known, liquid-liquid phase separation can occur either upon heating or cooling, corresponding to lower and upper critical solution phase boundaries, respectively. Likewise, self-assembly transitions from a monomeric state to an organized polymeric state can proceed either upon increasing or decreasing temperature, and the concentration dependent ordering temperature is correspondingly called the "floor" or "ceiling" temperature. Motivated by the fact that some phase separating systems exhibit closed loop phase boundaries with two critical points, the present paper analyzes self-assembly analogs of re-entrant phase separation, i.e., re-entrant self-assembly. In particular, re-entrant self-assembly transitions are demonstrated to arise in thermally activated equilibrium self-assembling systems, when thermal activation is more favorable than chain propagation, and in equilibrium self-assembly near an adsorbing boundary where strong competition exists between adsorption and self-assembly. Apparently, the competition between interactions or equilibria generally underlies re-entrant behavior in both liquid-liquid phase separation and self-assembly transitions.

INTEGRATION OF SHIP HULL ASSEMBLY SEQUENCE PLANNING, SCHEDULING AND BUDGETING

Directory of Open Access Journals (Sweden)

Remigiusz Romuald Iwańkowicz

2015-02-01

Full Text Available The specificity of the yard work requires the particularly careful treatment of the issues of scheduling and budgeting in the production planning processes. The article presents the method of analysis of the assembly sequence taking into account the duration of individual activities and the demand for resources. A method of the critical path and resource budgeting were used. Modelling of the assembly was performed using the acyclic graphs. It has been shown that the assembly sequences can have very different feasible budget regions. The proposed model is applied to the assembly processes of large-scale welded structures, including the hulls of ships. The presented computational examples have a simulation character. They show the usefulness of the model and the possibility to use it in a variety of analyses.

A new flooding correlation development and its critical heat flux predictions under low air-water flow conditions in Savannah River Site assembly channels

International Nuclear Information System (INIS)

Lee, S.Y.

1993-01-01

The upper limit to countercurrent flow, namely, flooding, is important to analyze the reactor coolability during an emergency cooling system (ECS) phase as a result of a large-break loss-of-coolant accident (LOCA) such as a double-ended guillotine break in the Savannah River Site (SRS) reactor system. During normal operation, the reactor coolant system utilizes downward flow through concentric heated tubes with ribs, which subdivided each annular channel into four subchannels. In this paper, a new flooding correlation has been developed based on the analytical models and literature data for adiabatic, steady-state, one-dimensional, air-water flow to predict flooding phenomenon in the SRS reactor assembly channel, which may have a counter-current air-water flow pattern during the ECS phase. In addition, the correlation was benchmarked against the experimental data conducted under the Oak Ridge National Laboratory multislit channel, which is close to the SRS assembly geometry. Furthermore, the correlation has also been used as a constitutive relationship in a new two-component two-phase thermal-hydraulics code FLOWTRAN-TF, which has been developed for a detailed analysis of SRS reactor assembly behavior during LOCA scenarios. Finally, the flooding correlation was applied to the predictions of critical heat flux, and the results were compared with the data taken by the SRS heat transfer laboratory under a single annular channel with ribs and a multiannular prototypic test rig

Assembly, alignment and test of the Transiting Exoplanet Survey Satellite (TESS) optical assemblies

Science.gov (United States)

Balonek, Gregory; Brown, Joshua J.; Andre, James E.; Chesbrough, Christian D.; Chrisp, Michael P.; Dalpiaz, Michael; Lennon, Joseph; Richards, B. C.; Clark, Kristin E.

2017-08-01

The Transiting Exoplanet Survey Satellite (TESS) will carry four visible waveband, seven-element, refractive F/1.4 lenses, each with a 34 degree diagonal field of view. This paper describes the methods used for the assembly, alignment and test of the four flight optical assemblies. Prior to commencing the build of the four flight optical assemblies, a Risk Reduction Unit (RRU) was successfully assembled and tested [1]. The lessons learned from the RRU were applied to the build of the flight assemblies. The main modifications to the flight assemblies include the inking of the third lens element stray light mitigation, tighter alignment tolerances, and diamond turning for critical mechanical surfaces. Each of the optical assemblies was tested interferometrically and measured with a low coherence distance measuring interferometer (DMI) to predict the optimal shim thickness between the lens assembly and detector before -75°C environmental testing. In addition to individual test data, environmental test results from prior assemblies allow for the exploration of marginal performance differences between each of the optical assemblies.

Method for calculating the critical heat flux in mixed rod assemblies based on the tables of crisis in bundles

International Nuclear Information System (INIS)

Bobkov, V.P.

2000-01-01

The method for calculating the critical heat flux in the mixed rod assemblies, for example RBMK, containing three-four angle and peripheral macrocells, is presented. The method is based on generalization of experimental data in form of tables for the rods beams. It is recommended for the areas of parameters both provided for by experimental data and for others, where the data are absent. The advantages of the table method as follows: it is acceptable within a wide range of parameters and provides for smooth description of dependence of critical heat fluxes on these parameters; it is characterized by clearness, high reliability and accuracy and is easy in application [ru

Theoretical analysis of time-dependent neutron spectra in bulk assemblies

International Nuclear Information System (INIS)

Akimoto, Tadashi; Ogawa, Yuichi; Togawa, Orihiko.

1988-01-01

Time-dependent neutron spectra in an iron assembly and in a graphite assembly are obtained with the one-dimensional S N calculation, in order an attempt to investigate the availability of these spectra to the benchmark test by the LINAC-TOF method for evaluation of nuclear data and numerical methods. The group constants are taken from the JAERI FAST SET Version 1, 2 and the ABBN SET. It was demonstrated by a sensitivity test that the time-dependent neutron spectra are sensitive to changes in the inelastic scattering cross section data in the iron assembly and to changes in the elastic scattering cross section data in the graphite assembly. Moreover, it is shown that the time-dependent spectra in the graphite assembly are sensitive to the group structure. Because some information about the neutron transport phenomena which has not been obtained in the stationary spectra is observed in the time-dependent spectra, the availability of the benchmark test based on the time-dependent spectra is indicated from the theoretical analysis. (author)

Sulfur activation at the Little Boy-Comet Critical Assembly: a replica of the Hiroshima bomb

International Nuclear Information System (INIS)

Kerr, G.D.; Emery, J.F.; Pace, J.V. III.

1985-04-01

Studies have been completed on the activation of sulfur by fast neutrons from the Little Boy-Comet Critical Assembly which replicates the general features of the Hiroshima bomb. The complex effects of the bomb's design and construction on leakage of sulfur-activation neutrons were investigated both experimentally and theoretically. Our sulfur activation studies were performed as part of a larger program to provide benchmark data for testing of methods used in recent source-term calculations for the Hiroshima bomb. Source neutrons capable of activating sulfur play an important role in determining neutron doses in Hiroshima at a kilometer or more from the point of explosion. 37 refs., 5 figs., 6 tabs

Experimental investigation of critical velocity in a parallel plate research reactor fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Castro, Alfredo J.A.; Scuro, Nikolas L.; Andrade, Delvonei A., E-mail: ajcastro@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNE-SP), Sao Paulo, SP (Brazil)

2017-07-01

The fuel elements of a MTR (Material Testing Reactor) type nuclear reactor are mostly composed of aluminum coated fuel plates containing the core of uranium silica (U{sub 3}Si{sub 2}) dispersed in an aluminum matrix. These plates have a thickness of the order of millimeters and are much longer in relation to their thickness. They are arranged in parallel in the assembly of the fuel element to form channels between them a few millimeters in thickness, through which there is a flow of the coolant. This configuration, combined with the need for a flow at high flow rates to ensure the cooling of the fuel element in operation, may create problems of mechanical failure of fuel plate due to the vibration induced by the flow in the channels. In the case of critical velocity excessive permanent deflections of the plates can cause blockage of the flow channel in the reactor core and lead to overheating in the plates. For this study an experimental bench capable of high volume flows and a test section that simulates a plate-like fuel element with three cooling channels were developed. The dimensions of the test section were based on the dimensions of the Fuel Element of the Brazilian Multipurpose Reactor (RMB), whose project is being coordinated by the National Commission of Nuclear Energy (CNEN). The experiments performed attained the objective of reaching Miller's critical velocity condition. The critical velocity was reached with 14.5 m/s leading to the consequent plastic deformation of the flow channel plates. (author)

Commissioning and start-up of RA-8 critical assembly

International Nuclear Information System (INIS)

Lorenzo, N. de; Diaz, C.; Facchini, G.; Fernandez, C.; Fittipaldi, A.; Juracich, R.; Levanon, I.; Manceda, J.; Martinez, J.; Mogdan, R.; Perez, J.; Scarnichia, E.; Blaumann, H.; Gennuso, G.; Scotti, G.

1999-01-01

The RA-8 critical assembly was designed as one of the experimental facilities for the CAREM Reactor Project. This paper describes the activities developed during the cold and hot commissioning, pointing out the difficulties and the solutions applied (some of them original ones). Moreover, this paper will show the main features of the newest nuclear installation of CNEA making a brief description of its characteristics. Among the special circumstances related to the commissioning that are described in the paper we can mention the following: 1. The facility shares the building with the Thermohydraulic Assay Laboratory (L.E.T.), another experimental facility of CAREM, and thus some shared systems have already been working for many years before this start up. Special procedures for these systems were designed to verify the proper functioning under the new requirements. 2. A new driving mechanism, based in hydraulic cylinders, was used to move the control rods. The criteria for acceptance and a validation of the procedure completeness have been carried out. 3. The implementation of a power measurement system based in neutron noise. 4. Measurement of Power Distribution using direct gamma counting from the fuel elements. 5. The commissioning was interrupted for a ten-month period because the personnel involved had to carry out the commissioning of the Egyptian Research Reactor 2. Also, the common activities during a commissioning are described, pointing out the major steps carried out and the results obtained. The following are examples of these activities: 1. Environmental dose survey (before fuel loading and during other stages). 2. Test of equipment and systems isolated from the rest of the plant. 3. Integrated system test (two or more systems working at the same time). 4. Start-up and power operation simulations before fuel loading. 5. Fuel loading strategy during the approximation to criticality by mass. 6. Modification of systems' components to improve the

MCNP and OMEGA criticality calculations

International Nuclear Information System (INIS)

Seifert, E.

1998-04-01

The reliability of OMEGA criticality calculations is shown by a comparison with calculations by the validated and widely used Monte Carlo code MCNP. The criticality of 16 assemblies with uranium as fissionable is calculated with the codes MCNP (Version 4A, ENDF/B-V cross sections), MCNP (Version 4B, ENDF/B-VI cross sections), and OMEGA. Identical calculation models are used for the three codes. The results are compared mutually and with the experimental criticality of the assemblies. (orig.)

Measurement and analysis on dynamic behaviour of parallel-plate assembly in nuclear reactors

International Nuclear Information System (INIS)

Chen Junjie; Guo Changqing; Zou Changchuan

1997-01-01

Measurement and analysis on dynamic behaviour of parallel-plate assembly in nuclear reactors have been explored. The electromagnetic method, a new method of measuring and analysing dynamic behaviour with the parallel-plate assembly as the structure of multi-parallel-beams joining with single-beam, has been presented. Theoretical analysis and computation results of dry-modal natural frequencies show good agreement with experimental measurements

Structural characteristics of a graphite moderated critical assembly for a Zero Power reactor at IEA (Brazil)

International Nuclear Information System (INIS)

Almeida Ferreira, A.C. de; Hukai, R.Y.

1975-01-01

The structural characteristics of a graphite moderated core of a critical assembly to be installed in the Zero Power Reactor of IEA have been defined. These characteristics are the graphite block dimensions, the number and dimensions of the holes in the graphite, the pitch, the dimensions of the sticks of fuel and graphite to be inserted in the holes, and the mechanical reproducibility of the system. The composition of the fuel and moderator sticks were also defined. The main boundary conditions were the range of the relation C/U and C/TH used in commercial HTGR and the neutronics homogeneity

New fuel vault criticality analysis at Chinshan nuclear power station with new approaches to improve the storage flexibility

International Nuclear Information System (INIS)

Huang, P. H.

2010-10-01

The Chinshan new fuel vault (NFV) consists of 13 fuel storage racks, each rack may store 10 fuel assemblies. Prior to 2008, the NFV had never been used and the practice by the Taiwan Power Company (TPC) was to temporarily store the fuel assemblies in the shipping containers after received, until the inspection work was performed shortly before the outage, and then assemblies were loaded directly into the spent fuel pool (SFP). Starting from 2009, this practice has been revised since the new fuel contract would only supply a small amount of containers for storage, and the SFP would lose full-core-off load capability soon; therefore, use of NFV to store fuel assemblies following inspection becomes extremely crucial. The original Chinshan NFV criticality analysis was performed for the initial fuel design. Although many new fuel designs had been used (e.g., Atrium-10 reported in PBNC-14), no reanalysis had been performed because it was not anticipated that NFV would be used. Therefore, TPC requested the vendor to perform the analysis for Atrium-10. Originally, the vendor estimated that number of assemblies allowed to be stored would be limited severely to about 60. To enhance storage flexibility, Tpc proposed some new approaches: 1) All assemblies are assumed in vendor's standard method to contain a single limiting lattice for entire fuel length, it is suggested that axially zoned limiting lattices be selected based on characteristics of reloads to be delivered, and this significantly improves flexibility. 2) The maximum k-effective equation used by vendor was corrected (manufacturing tolerances were conservatively mistreated). Also, the vendor typically used 0.95 k-effective as the criterion, it is suggested that NUREG-0800 requirement (≤0.98 for optimum moderation conditions) be applied. After several iterations, all the 130 locations are allowed to store fuel. The analysis report has been approved by the authority in June 2008. (Author)

New fuel vault criticality analysis at Chinshan nuclear power station with new approaches to improve the storage flexibility

Energy Technology Data Exchange (ETDEWEB)

Huang, P. H., E-mail: u808966@taipower.com.t [Taiwan Power Company, Department of Nuclear Generation, 242 Roosevelt Rd., Sec. 3, Taipei, Taiwan (China)

2010-10-15

The Chinshan new fuel vault (NFV) consists of 13 fuel storage racks, each rack may store 10 fuel assemblies. Prior to 2008, the NFV had never been used and the practice by the Taiwan Power Company (TPC) was to temporarily store the fuel assemblies in the shipping containers after received, until the inspection work was performed shortly before the outage, and then assemblies were loaded directly into the spent fuel pool (SFP). Starting from 2009, this practice has been revised since the new fuel contract would only supply a small amount of containers for storage, and the SFP would lose full-core-off load capability soon; therefore, use of NFV to store fuel assemblies following inspection becomes extremely crucial. The original Chinshan NFV criticality analysis was performed for the initial fuel design. Although many new fuel designs had been used (e.g., Atrium-10 reported in PBNC-14), no reanalysis had been performed because it was not anticipated that NFV would be used. Therefore, TPC requested the vendor to perform the analysis for Atrium-10. Originally, the vendor estimated that number of assemblies allowed to be stored would be limited severely to about 60. To enhance storage flexibility, Tpc proposed some new approaches: 1) All assemblies are assumed in vendor's standard method to contain a single limiting lattice for entire fuel length, it is suggested that axially zoned limiting lattices be selected based on characteristics of reloads to be delivered, and this significantly improves flexibility. 2) The maximum k-effective equation used by vendor was corrected (manufacturing tolerances were conservatively mistreated). Also, the vendor typically used 0.95 k-effective as the criterion, it is suggested that NUREG-0800 requirement ({<=}0.98 for optimum moderation conditions) be applied. After several iterations, all the 130 locations are allowed to store fuel. The analysis report has been approved by the authority in June 2008. (Author)

Criticality evaluation of long term for spent fuel, using Scale

International Nuclear Information System (INIS)

Esquivel E, J.; Vargas E, S.; Ramirez S, J. R.

2013-10-01

Once carried out the spent fuel discharge, of the reactor core, this continues generating decay heat and diverse fission products, reason why is important to store this fuel inside containers able to dissipate the heat generated by the isotopes decay of the fuel and to maintain the fuels arrangement in subcritical condition. This means that: is necessary to assure the sub-criticality of those fuel assemblies in the time. This work, presents a criticality evaluation of fuel assemblies type PWR in a storage generic container. For this purpose have been used two codes: GeeWiz, to carry out the geometric model of the container with the fuel assemblies, and Keno, with which, the criticality of the full container with fuel is determined until a 10 6 years period. These codes are part of the package Scale. The specifications for each one of the analyzed components are based on a Benchmark document of the Nea/OECD, of where, the results that reports are compared with the obtained results by the realized analysis. (Author)

The Texts of the Instruments connected with the Agency's Assistance to Mexico in Establishing a Sub-Critical Assembly Project

Energy Technology Data Exchange (ETDEWEB)

NONE

1967-10-25

The texts of the Supply Agreement between the Agency and the Governments of Mexico and the United States of America, and of the Project Agreement between the Agency and the Government of Mexico, connected with the Agency's assistance to the latter Government in establishing a sub-critical assembly project, are reproduced in this document for the information of all Members. Both Agreements entered into force on 23 August 1967.

The Texts of the Instruments connected with the Agency's Assistance to Mexico in Establishing a Sub-Critical Assembly Project

Energy Technology Data Exchange (ETDEWEB)

NONE

1966-07-07

The texts of the Supply Agreement between the Agency and the Governments of Mexico and the United States of America, and of the Project Agreement between the Agency and the Government of Mexico in connection with the Agency's assistance to that Government in establishing a sub-critical assembly project.. are reproduced in this document for the information of all Members. Both Agreements entered into force on 20 June 1966.

The Texts of the Instruments connected with the Agency's Assistance to Mexico in Establishing a Sub-Critical Assembly Project

International Nuclear Information System (INIS)

1966-01-01

The texts of the Supply Agreement between the Agency and the Governments of Mexico and the United States of America, and of the Project Agreement between the Agency and the Government of Mexico in connection with the Agency's assistance to that Government in establishing a sub-critical assembly project.. are reproduced in this document for the information of all Members. Both Agreements entered into force on 20 June 1966

The Texts of the Instruments connected with the Agency's Assistance to Mexico in Establishing a Sub-Critical Assembly Project

International Nuclear Information System (INIS)

1967-01-01

The texts of the Supply Agreement between the Agency and the Governments of Mexico and the United States of America, and of the Project Agreement between the Agency and the Government of Mexico, connected with the Agency's assistance to the latter Government in establishing a sub-critical assembly project, are reproduced in this document for the information of all Members. Both Agreements entered into force on 23 August 1967

Management of research reactor; dynamic characteristics analysis for reactor structures related with vibration of HANARO fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Ahn, Chang Kee; Shim, Joo Sup [Shinwa Technology Information, Seoul (Korea)

2001-04-01

The objective of this study is to deduce the dynamic correlation between the fuel assembly and the reactor structure. Dynamic characteristics analyses for reactor structure related with vibration of HANARO fuel assembly have been performed For the dynamic characteristic analysis, the in-air models of the round and hexagonal flow tubes, 18-element and 36-element fuel assemblies, and reactor structure were developed. By calculating the hydrodynamic mass and distributing it on the in-air models, the in-water models of the flow tubes, the fuel assemblies, and the reactor structure were developed. Then, modal analyses for developed in-air and in-water models have been performed. Especially, two 18-element fuel assemblies and three 36-element fuel assemblies were included in the in-water reactor models. For the verification of the modal analysis results, the natural frequencies and the mode shapes of the fuel assembly were compared with those obtained from the experiment. Finally the analysis results of the reactor structure were compared with them performed by AECL Based on the reactor model without PCS piping, the in-water reactor model including the fuel assemblies was developed, and its modal analysis was performed. The analysis results demonstrate that there are no resonance between the fuel assembly and the reactor structures. 26 refs., 419 figs., 85 tabs. (Author)

Preclosure Criticality Analysis Process Report

International Nuclear Information System (INIS)

Thomas, D.A.

1999-01-01

The design approach for criticality of the disposal container and waste package will be dictated by existing regulatory requirements. This conclusion is based on the fact that preclosure operations and facilities have significant similarities to existing facilities and operations currently regulated by the NRC. The major difference would be the use of a risk-informed approach with burnup credit. This approach could reduce licensing delays and costs of the repository. The probability of success for this proposed seamless licensing strategy is increased, since there is precedence of regulation (10 CFR Part 63 and NUREG 1520) and commercial precedence for allowing burnup credit at sites similar to Yucca Mountain during preclosure. While NUREG 1520 is not directly applicable to a facility for handling spent nuclear fuel, the risk-informed approach to criticality analysis in NUREG 1520 is considered indicative of how the NRC will approach risk-informed criticality analysis at spent fuel facilities in the future. The types of design basis events which must be considered during the criticality safety analysis portion of the Integrated Safety Analysis (ISA) are those events which result in unanticipated moderation, loss of neutron absorber, geometric changes in the critical system, or administrative errors in waste form placement (loading) of the disposal container. The specific events to be considered must be based on the review of the system's design, as discussed in Section 3.2. A transition of licensing approach (e.g., deterministic versus risk-informed, performance-based) is not obvious and will require analysis. For commercial spent nuclear fuel, the probability of interspersed moderation may be low enough to allow nearly the same Critical Limit for both preclosure and postclosure, though an administrative margin will be applied to preclosure and possibly not to postclosure. Similarly the Design Basis Events for the waste package may be incredible and therefore not

Preclosure Criticality Analysis Process Report

International Nuclear Information System (INIS)

Thomas, D.A.

1999-01-01

The design approach for criticality of the disposal container and waste package will be dictated by existing regulatory requirements. This conclusion is based on the fact that preclosure operations and facilities have significant similarities to existing facilities and operations currently regulated by the NRC. The major difference would be the use of a risk-informed approach with burnup credit. This approach could reduce licensing delays and costs of the repository. The probability of success for this proposed seamless licensing strategy is increased, since there is precedence of regulation (10 CFR Part 63 and NUREG 1520) and commercial precedence for allowing burnup credit at sites similar to Yucca Mountain during preclosure. While NUREG 1520 is not directly applicable to a facility for handling spent nuclear fuel, the risk-informed approach to criticality analysis in NUREG 1520 is considered indicative of how the NRC will approach risk-informed criticality analysis at spent fuel facilities in the future. The types of design basis events which must be considered during the criticality safety analysis portion of the Integrated Safety Analysis (ISA) are those events which result in unanticipated moderation, loss of neutron absorber, geometric changes in the critical system, or administrative errors in waste form placement (loading) of the disposal container. The specific events to be considered must be based on the review of the system's design, as discussed in Section 3.2. A transition of licensing approach (e.g., deterministic versus risk-informed, performance-based) is not obvious and will require analysis. For commercial spent nuclear fuel, the probability of interspersed moderation may be low enough to allow nearly the same Critical Limit for both preclosure and postclosure, though an administrative margin will be applied to preclosure and possibly not to postclosure. Similarly the Design Basis Events for the waste package may be incredible and therefore not

Analytical and experimental analysis of YALINA-Booster and YALINA-Thermal assemblies

International Nuclear Information System (INIS)

Kiyavitskaya, H.; Bournos, V.; Mazanik, S.; Khilmanovich, A.; Martsinkevich, B.; Routkovskaya, Ch.; Edchik, I.; Fokov, Y.; Sadovich, S.; Fedorenko, A.; Gohar, Y.; Talamo, A.

2010-01-01

Full text: Accelerator Driven Systems (ADS) may play an important role in future nuclear fuel cycles to reduce the longterm radiotoxicity and volume of spent nuclear fuel. It is proposed that ADS will produce energy and incinerate radioactive waste. This technology was called Accelerator Driven Transmutation Technology (ADTT). The most important problems of this technology are monitoring of a reactivity level in on-line regime, a choice of neutron spectrum appropriate for incineration of Minor Actinides (MA) and transmutation of Long Lived Fission Products (LLFP) and etc. Before the designing and construction of an installation it is necessary to carry out R and D to validate codes, nuclear data libraries and other instrumentations. The YALINA facility is designed to study the ADS physics and to investigate the transmutation reaction rates of MA and LLFP. The main objective of the YALINA benchmark is to compare the results from different calculation methods with each other and experimental data. The benchmark is based on the current YALINA facility configuration, which provides the opportunity to verify the prediction capability of the different methods. The experimental data have been obtained in the frame of the ISTC Projects B1341 'Analytical and experimental evaluation of the possibility to create a universal volume source of neutrons in the sub-critical booster assembly with low enrichment uranium fuel driven by a neutron generator' and B1732P 'Analytical and experimental evaluating the possibility of creation of universal volume source of neutrons in the sub-critical booster assembly with low enriched uranium fuel driven by the neutron generator'. In this paper a comparison of the experimental and calculated data obtained for YALINA-Booster subcritical assembly with a fuel of different enrichment and for YALINA-Thermal with a different number of control rods (216, 245 and 280) will be done.

Analysis of Bracket Assembly for Portable Leak Detector Station

International Nuclear Information System (INIS)

ZIADA, H.H.

1999-01-01

This Supporting Document Presents Structural and Stress Analysis of a Portable Leak Detector Station for Tank Farms. The results show that the bracket assembly meets the requirements for dead load and natural phenomena hazards loads (seismic and wind)

Neutron Activation Foil and Thermoluminescent Dosimeter Responses to a Lead Reflected Pulse of the CEA Valduc SILENE Critical Assembly

International Nuclear Information System (INIS)

Miller, Thomas Martin; Celik, Cihangir; Isbell, Kimberly McMahan; Lee, Yi-kang; Gagnier, Emmanuel; Authier, Nicolas; Piot, Jerome; Jacquet, Xavier; Rousseau, Guillaume; Reynolds, Kevin H.

2016-01-01

This benchmark experiment was conducted as a joint venture between the US Department of Energy (DOE) and the French Commissariat à l'Energie Atomique (CEA). Staff at the Oak Ridge National Laboratory (ORNL) in the US and the Centre de Valduc in France planned this experiment. The experiment was conducted on October 13, 2010 in the SILENE critical assembly facility at Valduc. Several other organizations contributed to this experiment and the subsequent evaluation, including CEA Saclay, Lawrence Livermore National Laboratory (LLNL), the Y-12 National Security Complex (NSC), Babcock International Group in the United Kingdom, and Los Alamos National Laboratory (LANL). The goal of this experiment was to measure neutron activation and thermoluminescent dosimeter (TLD) doses from a source similar to a fissile solution critical excursion. The resulting benchmark can be used for validation of computer codes and nuclear data libraries as required when performing analysis of criticality accident alarm systems (CAASs). A secondary goal of this experiment was to qualitatively test performance of two CAAS detectors similar to those currently and formerly in use in some US DOE facilities. The detectors tested were the CIDAS MkX and the Rocky Flats NCD-91. The CIDAS detects gammas with a Geiger-Muller tube, and the Rocky Flats detects neutrons via charged particles produced in a thin 6 LiF disc, depositing energy in a Si solid-state detector. These detectors were being evaluated to determine whether they would alarm, so they were not expected to generate benchmark quality data.

Neutron Activation Foil and Thermoluminescent Dosimeter Responses to a Polyethylene Reflected Pulse of the CEA Valduc SILENE Critical Assembly

International Nuclear Information System (INIS)

Miller, Thomas Martin; Celik, Cihangir; McMahan, Kimberly L.; Lee, Yi-kang; Gagnier, Emmanuel; Authier, Nicolas

2016-01-01

This benchmark experiment was conducted as a joint venture between the US Department of Energy (DOE) and the French Commissariat à l'Energie Atomique (CEA). Staff at the Oak Ridge National Laboratory (ORNL) in the US and the Centre de Valduc in France planned this experiment. The experiment was conducted on October 19, 2010 in the SILENE critical assembly facility at Valduc. Several other organizations contributed to this experiment and the subsequent evaluation, including CEA Saclay, Lawrence Livermore National Laboratory (LLNL), the Y-12 National Security Complex (NSC), Babcock International Group in the United Kingdom, and Los Alamos National Laboratory (LANL). The goal of this experiment was to measure neutron activation and thermoluminescent dosimeter (TLD) doses from a source similar to a fissile solution critical excursion. The resulting benchmark can be used for validation of computer codes and nuclear data libraries as required when performing analysis of criticality accident alarm systems (CAASs). A secondary goal of this experiment was to qualitatively test performance of two CAAS detectors similar to those currently and formerly in use in some US DOE facilities. The detectors tested were the CIDAS MkX and the Rocky Flats NCD-91. The CIDAS detects gammas with a Geiger-Muller tube and the Rocky Flats detects neutrons via charged particles produced in a thin "6LiF disc depositing energy in a Si solid state detector. These detectors were being evaluated to determine whether they would alarm, so they were not expected to generate benchmark quality data.

Neutron Activation Foil and Thermoluminescent Dosimeter Responses to a Polyethylene Reflected Pulse of the CEA Valduc SILENE Critical Assembly

Energy Technology Data Exchange (ETDEWEB)

Miller, Thomas Martin [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Celik, Cihangir [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McMahan, Kimberly L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lee, Yi-kang [French Atomic Energy Commission (CEA), Saclay (France); Gagnier, Emmanuel [French Atomic Energy Commission (CEA), Centre de Saclay, Gif sur Yvette; Authier, Nicolas [French Atomic Energy Commission (CEA), Salives (France). Valduc Centre for Nuclear Studies; Piot, Jerome [French Atomic Energy Commission (CEA), Salives (France). Valduc Centre for Nuclear Studies; Jacquet, Xavier [French Atomic Energy Commission (CEA), Salives (France). Valduc Centre for Nuclear Studies; Rousseau, Guillaume [French Atomic Energy Commission (CEA), Salives (France). Valduc Centre for Nuclear Studies; Reynolds, Kevin H. [Y-12 National Security Complex, Oak Ridge, TN (United States)

2016-09-01

This benchmark experiment was conducted as a joint venture between the US Department of Energy (DOE) and the French Commissariat à l'Energie Atomique (CEA). Staff at the Oak Ridge National Laboratory (ORNL) in the US and the Centre de Valduc in France planned this experiment. The experiment was conducted on October 19, 2010 in the SILENE critical assembly facility at Valduc. Several other organizations contributed to this experiment and the subsequent evaluation, including CEA Saclay, Lawrence Livermore National Laboratory (LLNL), the Y-12 National Security Complex (NSC), Babcock International Group in the United Kingdom, and Los Alamos National Laboratory (LANL). The goal of this experiment was to measure neutron activation and thermoluminescent dosimeter (TLD) doses from a source similar to a fissile solution critical excursion. The resulting benchmark can be used for validation of computer codes and nuclear data libraries as required when performing analysis of criticality accident alarm systems (CAASs). A secondary goal of this experiment was to qualitatively test performance of two CAAS detectors similar to those currently and formerly in use in some US DOE facilities. The detectors tested were the CIDAS MkX and the Rocky Flats NCD-91. The CIDAS detects gammas with a Geiger-Muller tube and the Rocky Flats detects neutrons via charged particles produced in a thin ⁶LiF disc depositing energy in a Si solid state detector. These detectors were being evaluated to determine whether they would alarm, so they were not expected to generate benchmark quality data.

Neutron Activation Foil and Thermoluminescent Dosimeter Responses to a Lead Reflected Pulse of the CEA Valduc SILENE Critical Assembly

Energy Technology Data Exchange (ETDEWEB)

Miller, Thomas Martin [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Celik, Cihangir [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Isbell, Kimberly McMahan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lee, Yi-kang [Commissariat a l' Energie Atomique et aux Energies Alternatives (CEA-Saclay), Gif-sur-Yvette (France); Gagnier, Emmanuel [Commissariat a l' Energie Atomique et aux Energies Alternatives (CEA-Saclay), Gif-sur-Yvette (France); Authier, Nicolas [Commissariat a l' Energie Atomique et aux Energies Alternatives (CEA-Saclay), Gif-sur-Yvette (France); Piot, Jerome [Commissariat a l' Energie Atomique et aux Energies Alternatives (CEA-Saclay), Gif-sur-Yvette (France); Jacquet, Xavier [Commissariat a l' Energie Atomique et aux Energies Alternatives (CEA-Saclay), Gif-sur-Yvette (France); Rousseau, Guillaume [Commissariat a l' Energie Atomique et aux Energies Alternatives (CEA-Saclay), Gif-sur-Yvette (France); Reynolds, Kevin H. [Y-12 National Security Complex, Oak Ridge, TN (United States)

2016-09-01

This benchmark experiment was conducted as a joint venture between the US Department of Energy (DOE) and the French Commissariat à l'Energie Atomique (CEA). Staff at the Oak Ridge National Laboratory (ORNL) in the US and the Centre de Valduc in France planned this experiment. The experiment was conducted on October 13, 2010 in the SILENE critical assembly facility at Valduc. Several other organizations contributed to this experiment and the subsequent evaluation, including CEA Saclay, Lawrence Livermore National Laboratory (LLNL), the Y-12 National Security Complex (NSC), Babcock International Group in the United Kingdom, and Los Alamos National Laboratory (LANL). The goal of this experiment was to measure neutron activation and thermoluminescent dosimeter (TLD) doses from a source similar to a fissile solution critical excursion. The resulting benchmark can be used for validation of computer codes and nuclear data libraries as required when performing analysis of criticality accident alarm systems (CAASs). A secondary goal of this experiment was to qualitatively test performance of two CAAS detectors similar to those currently and formerly in use in some US DOE facilities. The detectors tested were the CIDAS MkX and the Rocky Flats NCD-91. The CIDAS detects gammas with a Geiger-Muller tube, and the Rocky Flats detects neutrons via charged particles produced in a thin ⁶LiF disc, depositing energy in a Si solid-state detector. These detectors were being evaluated to determine whether they would alarm, so they were not expected to generate benchmark quality data.

Impact analysis of spent fuel jacket assemblies

International Nuclear Information System (INIS)

Aramayo, G.A.

1994-01-01

As part of the analyses performed in support of the reracking of the High Flux Isotope Reactor pool, it became necessary to prove the structural integrity of the spent fuel jacket assemblies subjected to gravity drop that result from postulated accidents associated with the handling of these assemblies while submerged in the pool. The spent fuel jacket assemblies are an integral part of the reracking project, and serve to house fuel assemblies. The structure integrity of the jacket assemblies from loads that result from impact from a height of 10 feet onto specified targets has been performed analytically using the computer program LS-DYNA3D. Nine attitudes of the assembly at the time of impact have been considered. Results of the analyses show that there is no failure of the assemblies as a result of the impact scenarios considered

The single SNR fuel assembly container (ESBB) to transport unirradiated SNR 300 fuel assemblies

International Nuclear Information System (INIS)

Hilbert, F.; Hottenrott, G.

1998-01-01

In this paper a new type B(U) package design is presented. The Single SNR Fuel Assembly Container (ESBB) is designed for the transport and storage of a single SNR 300 fuel assembly. This package is the main component for the future interim storage of the fuel assemblies in heavy storage casks. Its benefits are that it is compatible with the Category I transport system of Nuclear Cargo + Service NCS) used in Germany and that it can be easily handled at the current storage locations as well as in an interim storage facility. In total 205 fuel assemblies are currently stored in Hanau, Germany and Dounreay, U.K. Former studies have shown, that heavy transport and storage casks can be handled there only with considerable efforts. But the required category I transport to an interim storage is not reasonably feasible. To overcome these problems the ESBB was designed. It consists of a stainless steel tube with welded bottom, a welded plug as closure system and shock absorbers 26 packages at maximum can be transported in one batch with the NCS security vehicle. The safety analysis shows that the package complies with IAEA 1996. Standard calculations methods and computer codes like HEATING 7.2 (Childs 1993) have been used for the analysis. Criticality safety assessment is based on conservative assumptions as required in IAEA 1996. Drop tests carried out by BAM will be used to verify the design. These tests are scheduled for mid 1998. For the validation of the design prototypes have already been manufactured. Handling tests show that the design complies with the requirements. Preliminary drop tests show that the certification drop tests will be passed positively. (authors)

History of critical experiments at Pajarito Site

International Nuclear Information System (INIS)

Paxton, H.C.

1983-03-01

This account describes critical and subcritical assemblies operated remotely at the Pajarito Canyon Site at the Los Alamos National Laboratory. Earliest assemblies, directed exclusively toward the nuclear weapons program, were for safety tests. Other weapon-related assemblies provided neutronic information to check detailed weapon calculations. Topsy, the first of these critical assemblies, was followed by Lady Godiva, Jezebel, Flattop, and ultimately Big Ten. As reactor programs came to Los Alamos, design studies and mockups were tested at Pajarito Site. For example, nearly all 16 Rover reactors intended for Nevada tests were preceded by zero-power mockups and proof tests at Pajarito Site. Expanded interest and capability led to fast-pulse assemblies, culminating in Godiva IV and Skua, and to the Kinglet and Sheba solution assemblies

History of critical experiments at Pajarito Site

Energy Technology Data Exchange (ETDEWEB)

Paxton, H.C.

1983-03-01

This account describes critical and subcritical assemblies operated remotely at the Pajarito Canyon Site at the Los Alamos National Laboratory. Earliest assemblies, directed exclusively toward the nuclear weapons program, were for safety tests. Other weapon-related assemblies provided neutronic information to check detailed weapon calculations. Topsy, the first of these critical assemblies, was followed by Lady Godiva, Jezebel, Flattop, and ultimately Big Ten. As reactor programs came to Los Alamos, design studies and mockups were tested at Pajarito Site. For example, nearly all 16 Rover reactors intended for Nevada tests were preceded by zero-power mockups and proof tests at Pajarito Site. Expanded interest and capability led to fast-pulse assemblies, culminating in Godiva IV and Skua, and to the Kinglet and Sheba solution assemblies.

Investigation of the neutron detection statistics in fast critical assembly BFS-24-1

International Nuclear Information System (INIS)

Avramov, A.M.; Tyutyunnikov, P.L.; Mikulski, A.T.; Rafalska, E.; Chwaszczewski, S.; Jablonski, K.

1974-01-01

The results of the neutron detection statistics investigation at the fast critical assembly BFS-24-1 are given. The Ross-α measurements were carried out using: digital flash-start unit and 256 channel time analyzer, 10 channel time analyzer, alphameter device. Parallely the measurements using the variable dead time method and zero probability method were performed. The prompt neutron decay constants, the effectiveness of neutron detector and the intensity of external neutron source are determined using the experimental data. The experimental values of prompt neutron decay constant are compared with the calculated ones. The codes used in the calculation are following: one dimensional, diffusion, 26-group code 26-M and EWA-1, one dimensional, multiregion, nonstationary diffusion 3-group code SPECTR, 26-group, diffusion code in buckling approximation, MIXSPECTR. In all codes the 26 group nuclear constants BNAB-26 and BNAB-70 are used. (author)

Critical experiments with mixed oxide fuel

International Nuclear Information System (INIS)

Harris, D.R.

1997-01-01

This paper very briefly outlines technical considerations in performing critical experiments on weapons-grade plutonium mixed oxide fuel assemblies. The experiments proposed would use weapons-grade plutonium and Er 2 O 3 at various dissolved boron levels, and for specific fuel assemblies such as the ABBCE fuel assembly with five large water holes. Technical considerations described include the core, the measurements, safety, security, radiological matters, and licensing. It is concluded that the experiments are feasible at the Rensselaer Polytechnic Institute Reactor Critical Facility. 9 refs

Rapid centriole assembly in Naegleria reveals conserved roles for both de novo and mentored assembly.

Science.gov (United States)

Fritz-Laylin, Lillian K; Levy, Yaron Y; Levitan, Edward; Chen, Sean; Cande, W Zacheus; Lai, Elaine Y; Fulton, Chandler

2016-03-01

Centrioles are eukaryotic organelles whose number and position are critical for cilia formation and mitosis. Many cell types assemble new centrioles next to existing ones ("templated" or mentored assembly). Under certain conditions, centrioles also form without pre-existing centrioles (de novo). The synchronous differentiation of Naegleria amoebae to flagellates represents a unique opportunity to study centriole assembly, as nearly 100% of the population transitions from having no centrioles to having two within minutes. Here, we find that Naegleria forms its first centriole de novo, immediately followed by mentored assembly of the second. We also find both de novo and mentored assembly distributed among all major eukaryote lineages. We therefore propose that both modes are ancestral and have been conserved because they serve complementary roles, with de novo assembly as the default when no pre-existing centriole is available, and mentored assembly allowing precise regulation of number, timing, and location of centriole assembly. © 2016 Wiley Periodicals, Inc.

The analysis of FCA critical experiments and its application to ''JOYO'' nuclear design

International Nuclear Information System (INIS)

Iijima, S.

1979-01-01

A series of extensive mockup experiments in support of Japanese Experimental Fast Reactor, ''JOYO'', were performed at Fast Critical Assembly in JAERI, from February 1970 to March 1972. The present paper describes the results of analysis of these mockup experiments and its application to ''JOYO'' nuclear design. The basic calculational method of the analysis is the same as that employed in ''JOYO'' neutronics calculation, viz., the 6-group diffusion theory using 25-group NAIG Nuclear Set No. 5. Corrections to the base calculations were evaluated by using one-dimensional S 4 transport theory and integral transport theory. The ABBN group constants were also used for the sake of comparison. The most probable values of JOYO neutronics parameters were determined by applying the bias factor (E/C) to the calculated values. The uncertainties of the most probable values were also determined, and they were taken into consideration in the JOYO design

ANALYSIS OF THE GAZE BEHAVIOUR OF THE WORKER ON THE CARBURETOR ASSEMBLY TASK

Directory of Open Access Journals (Sweden)

Novie Susanto

2015-06-01

Full Text Available This study presents analysis of the area of interest (AOI and the gaze behavior of human during assembly task. This study aims at investigating the human behavior in detail using an eye‐tracking system during assembly task using LEGO brick and an actual manufactured product, a carburetor. An analysis using heat map data based on the recorded videos from the eye-tracking system is taken into account to examine and investigate the gaze behavior of human. The results of this study show that the carburetor assembly requires more attention than the product made from LEGO bricks. About 50% of the participants experience the necessity to visually inspect the interim state of the work object during the simulation of the assembly sequence on the screen. They also show the tendency to want to be more certain about part fitting in the actual work object.

A real time analysis of the self-assembly process using thermal analysis inside the differential scanning calorimeter instrument.

Science.gov (United States)

Roy, Debmalya; Shastri, Babita; Mukhopadhyay, K

2012-07-12

The supramolecular assembly of the regioregular poly-3-hexylthiophene (rr-P3HT) in solution has been investigated thoroughly in the past. In the current study, our focus is on the enthalpy of nanofiber formation using thermal analysis techniques by performing the self-assembly process inside the differential scanning calorimetry (DSC) instrument. Thermogravimetric analysis (TGA) was carried out to check the concentration of the solvent during the self-assembly process of P3HT in p-xylene. Ultraviolet visible (UV-vis) spectophotometric technique, small-angle X-ray scattering (SAXS) experiment, atomic force microscopic (AFM), and scanning electron microscopic (SEM) images were used to characterize the different experimental yields generated by cooling the reaction mixture at desired temperatures. Comparison of the morphologies of self-assembled products at different fiber formation temperatures gives us an idea about the possible crystallization parameters which could affect the P3HT nanofiber morphology.

Design analysis of a new SCWR fuel assembly using a coupled method

International Nuclear Information System (INIS)

Liu Xiaojing; Yang Ting; Cheng Xu

2011-01-01

Among the six GEN-Ⅳ reactor concepts recommended by the Gen-Ⅳ International Forum (GIF), supercritical water-cooled reactor (SCWR) is the only reactor type with water as coolant. Compared to the existing reactors, it has economic advantage and technology continuity. Based on the newly developed coupling code, analysis on the square SCWR assembly is carried out in this paper. A new design concept of SCWR fuel assembly is proposed. The results achieved so far indicate favorable thermal-hydraulic performance and neutron-physical behavior of the new fuel assembly compared to the previous ones. (authors)

ACRR fuel storage racks criticality safety analysis

International Nuclear Information System (INIS)

Bodette, D.E.; Naegeli, R.E.

1997-10-01

This document presents the criticality safety analysis for a new fuel storage rack to support modification of the Annular Core Research Reactor for production of molybdenum-99 at Sandia National Laboratories, Technical Area V facilities. Criticality calculations with the MCNP code investigated various contingencies for the criticality control parameters. Important contingencies included mix of fuel element types stored, water density due to air bubbles or water level for the over-moderated racks, interaction with existing fuel storage racks and fuel storage holsters in the fuel storage pool, neutron absorption of planned rack design and materials, and criticality changes due to manufacturing tolerances or damage. Some limitations or restrictions on use of the new fuel storage rack for storage operations were developed through the criticality analysis and are required to meet the double contingency requirements of criticality safety. As shown in the analysis, this system will remain subcritical under all credible upset conditions. Administrative controls are necessary for loading, moving, and handling the storage rack as well as for control of operations around it. 21 refs., 16 figs., 4 tabs

Criticality safety

International Nuclear Information System (INIS)

Walker, G.

1983-01-01

When a sufficient quantity of fissile material is brought together a self-sustaining neutron chain reaction will be started in it and will continue until some change occurs in the fissile material to stop the chain reaction. The quantity of fissile material required is the 'Critical Mass'. This is not a fixed quantity even for a given type of fissile material but varies between quite wide limits depending on a number of factors. In a nuclear reactor the critical mass of fissile material is assembled under well-defined condition to produce a controllable chain reaction. The same materials have to be handled outside the reactor in all stages of fuel element manufacture, storage, transport and irradiated fuel reprocessing. At any stage it is possible (at least in principle) to assemble a critical mass and thus initiate an accidental and uncontrollable chain reaction. Avoiding this is what criticality safety is all about. A system is just critical when the rate of production of neutrons balances the rate of loss either by escape or by absorption. The factors affecting criticality are, therefore, those which effect neutron production and loss. The principal ones are:- type of nuclide and enrichment (or isotopic composition), moderation, reflection, concentration (density), shape and interaction. Each factor is considered in detail. (author)

Vibration analysis of reactor assembly internals for Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Chellapandi, P.; Jalaldeen, S.; Srinivasan, R.; Chetal, S.C.; Bhoje, S.B.

2003-01-01

Vibration analysis of the reactor assembly components of 500 MWe Prototype Fast Breeder Reactor (PFBR) is presented. The vibration response of primary pump as well as dynamic forces developed at its supports are predicted numerically. The stiffness properties of hydrostatic bearing are determined by formulating and solving governing fluid and structural mechanics equations. The dynamic forces exerted by pump are used as input data for the dynamic response of reactor assembly components, mainly inner vessel, thermal baffle and control plug. Dynamic response of reactor assembly components is also predicted for the pressure fluctuations caused by sodium free level oscillations. Thermal baffle (weir shell) which is subjected to fluid forces developed at the associated sodium free levels is analysed by formulating and solving a set of non-linear equations for fluids, structures and fluid structure interaction (FSI). The control rod drive mechanism is analysed for response under flow induced forces on the parts subjected to cross flow in the zone just above the core top, taking into account FSI between sheaths of control and safety rod and absorber pin bundle. Based on the analysis results, it is concluded that the reactor assembly internals are free from any risk of mechanical as well as flow induced vibrations. (author)

Development of 3-D Flow Analysis Code for Fuel Assembly using Unstructured Grid System

Energy Technology Data Exchange (ETDEWEB)

Myong, Hyon Kook; Kim, Jong Eun; Ahn, Jong Ki; Yang, Seung Yong [Kookmin Univ., Seoul (Korea, Republic of)

2007-03-15

The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Final objective of this study is to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system. In order to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system, the following researches are made: - Development of numerical algorithm for CFD code's solver - Grid and geometric connectivity data - Development of software(PowerCFD code) for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system - Modulation of software(PowerCFD code) - Development of turbulence model - Development of analysis module of RANS/LES hybrid models - Analysis of turbulent flow and heat transfer - Basic study on LES analysis - Development of main frame on pre/post processors based on GUI - Algorithm for fully-developed flow.

Development of 3-D Flow Analysis Code for Fuel Assembly using Unstructured Grid System

International Nuclear Information System (INIS)

Myong, Hyon Kook; Kim, Jong Eun; Ahn, Jong Ki; Yang, Seung Yong

2007-03-01

The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Final objective of this study is to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system. In order to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system, the following researches are made: - Development of numerical algorithm for CFD code's solver - Grid and geometric connectivity data - Development of software(PowerCFD code) for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system - Modulation of software(PowerCFD code) - Development of turbulence model - Development of analysis module of RANS/LES hybrid models - Analysis of turbulent flow and heat transfer - Basic study on LES analysis - Development of main frame on pre/post processors based on GUI - Algorithm for fully-developed flow

Spectral measurements in critical assemblies: MCNP specifications and calculated results

Energy Technology Data Exchange (ETDEWEB)

Stephanie C. Frankle; Judith F. Briesmeister

1999-12-01

Recently, a suite of 86 criticality benchmarks for the Monte Carlo N-Particle (MCNP) transport code was developed, and the results of testing the ENDF/B-V and ENDF/B-VI data (through Release 2) were published. In addition to the standard k{sub eff} measurements, other experimental measurements were performed on a number of these benchmark assemblies. In particular, the Cross Section Evaluation Working Group (CSEWG) specifications contain experimental data for neutron leakage and central-flux measurements, central-fission ratio measurements, and activation ratio measurements. Additionally, there exists another set of fission reaction-rate measurements performed at the National Institute of Standards and Technology (NIST) utilizing a {sup 252}Cf source. This report will describe the leakage and central-flux measurements and show a comparison of experimental data to MCNP simulations performed using the ENDF/B-V and B-VI (Release 2) data libraries. Central-fission and activation reaction-rate measurements will be described, and the comparison of experimental data to MCNP simulations using available data libraries for each reaction of interest will be presented. Finally, the NIST fission reaction-rate measurements will be described. A comparison of MCNP results published previously with the current MCNP simulations will be presented for the NIST measurements, and a comparison of the current MCNP simulations to the experimental measurements will be presented.

Spectral measurements in critical assemblies: MCNP specifications and calculated results

International Nuclear Information System (INIS)

Frankle, Stephanie C.; Briesmeister, Judith F.

1999-01-01

Recently, a suite of 86 criticality benchmarks for the Monte Carlo N-Particle (MCNP) transport code was developed, and the results of testing the ENDF/B-V and ENDF/B-VI data (through Release 2) were published. In addition to the standard k eff measurements, other experimental measurements were performed on a number of these benchmark assemblies. In particular, the Cross Section Evaluation Working Group (CSEWG) specifications contain experimental data for neutron leakage and central-flux measurements, central-fission ratio measurements, and activation ratio measurements. Additionally, there exists another set of fission reaction-rate measurements performed at the National Institute of Standards and Technology (NIST) utilizing a 252 Cf source. This report will describe the leakage and central-flux measurements and show a comparison of experimental data to MCNP simulations performed using the ENDF/B-V and B-VI (Release 2) data libraries. Central-fission and activation reaction-rate measurements will be described, and the comparison of experimental data to MCNP simulations using available data libraries for each reaction of interest will be presented. Finally, the NIST fission reaction-rate measurements will be described. A comparison of MCNP results published previously with the current MCNP simulations will be presented for the NIST measurements, and a comparison of the current MCNP simulations to the experimental measurements will be presented

A static analysis tool set for assembler code verification

International Nuclear Information System (INIS)

Dhodapkar, S.D.; Bhattacharjee, A.K.; Sen, Gopa

1991-01-01

Software Verification and Validation (V and V) is an important step in assuring reliability and quality of the software. The verification of program source code forms an important part of the overall V and V activity. The static analysis tools described here are useful in verification of assembler code. The tool set consists of static analysers for Intel 8086 and Motorola 68000 assembly language programs. The analysers examine the program source code and generate information about control flow within the program modules, unreachable code, well-formation of modules, call dependency between modules etc. The analysis of loops detects unstructured loops and syntactically infinite loops. Software metrics relating to size and structural complexity are also computed. This report describes the salient features of the design, implementation and the user interface of the tool set. The outputs generated by the analyser are explained using examples taken from some projects analysed by this tool set. (author). 7 refs., 17 figs

Analysis of the thermomechanical behavior of the IFMIF bayonet target assembly under design loading scenarios

Energy Technology Data Exchange (ETDEWEB)

Bernardi, D., E-mail: davide.bernardi@enea.it [ENEA Brasimone, Camugnano, BO (Italy); Arena, P.; Bongiovì, G.; Di Maio, P.A. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, Palermo (Italy); Frisoni, M. [ENEA Bologna, Via Martiri di Monte Sole 4, Bologna (Italy); Miccichè, G.; Serra, M. [ENEA Brasimone, Camugnano, BO (Italy)

2015-10-15

In the framework of the IFMIF Engineering Validation and Engineering Design Activities (IFMIF/EVEDA) phase, ENEA is responsible for the design of the European concept of the IFMIF lithium target system which foresees the possibility to periodically replace only the most irradiated and thus critical component (i.e., the backplate) while continuing to operate the rest of the target for a longer period (the so-called bayonet backplate concept). In this work, the results of the steady state thermomechanical analysis of the IFMIF bayonet target assembly under two different design loading scenarios (a “hot” scenario and a “cold” scenario) are briefly reported highlighting the relevant indications obtained with respect to the fulfillment of the design requirements. In particular, the analyses have shown that in the hot scenario the temperatures reached in the target assembly are within the material acceptable limits while in the cold scenario transition below the ductile to brittle transition temperature (DBTT) cannot be excluded. Moreover, results indicate that the contact between backplate and high flux test module is avoided and that the overall structural integrity of the system is assured in both scenarios. However, stress linearization analysis reveals that ITER Structural Design Criteria for In-vessel Components (SDC-IC) design rules are not always met along the selected paths at backplate middle plane section in the hot scenario, thus suggesting the need of a revision of the backplate design or a change of the operating conditions.

Nuclear safety analysis for transport cask TK-6 (for WWER-440) and cover for fresh assemblies (for WWER-1000) in implementation of new fuel types at Ukrainian NPP

International Nuclear Information System (INIS)

Bilodid, Y.; Kovbasenko, Iu; Dudka, Olena

2006-01-01

According to the fresh fuel management procedure, fuel assemblies - after nuclear fuel delivery to the NPP fresh fuel unit - are vertically loaded into a cover intended for the delivery of fuel assemblies into the containment of the NPP reactor compartment. The cover is placed into an universal jack in the cooling and refueling pond, and then the fresh fuel assemblies are loaded into the reactor core. Based on the nuclear safety analysis carried out by the Russian Research Center 'Kurchatov Institute' for contemporary WWER-1000 fuel, it has become necessary to limit the number of fuel assemblies loaded into a cover below its designed capacity (12 FA instead of 18 FA as originally designed). Such a decision leads to worse economic performances in fuel transportation. The paper considers potential ways to overcome this restriction. Transport container TK-6 for spent fuel assemblies was designed quite a long time ago and, as shown in this paper, the requirement on the maximally permissible neutron multiplication factor of the loaded container for individual states to be analyzed in compliance with Ukrainian regulations is not met. First of all, this concerns the container criticality analysis in optimal neutron slow-down (container filling with water-air mixture with optimal density). The paper shows potential ways for TK-6 burnup-credit loading with the maximum number of fuel assemblies and partial container loading (Authors)

Stress corrosion cracking life estimation of hold-down spring screw for nuclear fuel assembly

International Nuclear Information System (INIS)

Koh, S.K.

2005-01-01

Hold-down spring screw fractures due to primary water stress corrosion cracking were observed in nuclear fuel assemblies. The screw fastens hold-down springs that are required to maintain the nuclear fuel assembly in contact with upper core plate and permit thermal and irradiation-induced length changes. In order to investigate the primary causes of the screw fractures, the finite element stress analysis and fracture mechanics analysis were performed on the hold-down spring assembly. The elastic-plastic finite element analysis showed that the local stresses at the critical regions of head-shank fillet and thread root significantly exceeded the yield strength of the screw material, resulting in local plastic deformation. Preloading on the screw applied for tightening had beneficial effects on the screw strength by reducing the stress level at the critical regions, compared to the screw without preload. Calculated deflections and strains at the hold-down springs using the finite element analysis were in very close agreements with the experimentally measured deflections and strains. Primary water stress corrosion cracking (PWSCC) life of the Inconel 600 screw was predicted by integrating the Scott's model and resulted in a life of 1.42 years, which was fairly close to the field experience. Cracks were expected to originate at the threaded region of the screw and propagated to the opposite side of the spring, which was confirmed by the fractographic analysis of the fractured screws. (orig.)

Storage arrangement for nuclear reactor fuel assemblies

International Nuclear Information System (INIS)

Wade, E.E.

1977-01-01

Said invention is intended for providing an arrangement of spent fuel assembly storage inside which the space is efficiently used without accumulating a critical mass. The storage is provided for long fuel assemblies having along their longitudinal axis an active part containing the fuel and an inactive part empty of fuel. Said storage arrangement comprises a framework constituting some long-shaped cells designed so as each of them can receive a fuel assembly. Means of axial positioning of said assembly in a cell make it possible to support the fuel assemblies inside the framework according to a spacing ratio, along the cell axis, such as the active part of an assembly is adjacent to the inactive part of the adjacent assemblies [fr

A CFD analysis of flow blockage phenomena in ALFRED LFR demo fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Di Piazza, Ivan, E-mail: ivan.dipiazza@enea.it [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone (Italy); Magugliani, Fabrizio [Ansaldo Nucleare, ANN, Corso Perrone n.25, Genova (Italy); Tarantino, Mariano [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone (Italy); Alemberti, Alessandro [Ansaldo Nucleare, ANN, Corso Perrone n.25, Genova (Italy)

2014-09-15

Highlights: • URANS simulations were performed on internal flow blockage in HLM fuel assemblies. • Comparison with RELAP results for foot blockage shows a very good agreement. • The temperature peak behind the blockage is dominant for large blockages. • A blockage of ∼15% leads to a maximum clad temperature around 800 °C in 3–4 s. • Local clad temperatures around 1000 °C are reached for blockages of 30% or more. - Abstract: A CFD study was carried out on fluid flow and heat transfer in the Lead-cooled Fuel Pin Bundle of the ALFRED LFR DEMO. In the context of GEN-IV heavy liquid metal-cooled reactors safety studies, the flow blockage in a fuel sub-assembly is considered one of the main issues to be addressed and the most important and realistic accident for LFR fuel assembly. The present paper is a first step toward a detailed analysis of such phenomena, and a CFD model and approach are presented to have a detailed thermo-fluid dynamic picture in the case of blockage. In particular the closed hexagonal, grid-spaced fuel assembly of the LFR ALFRED was modeled and computed. At this stage, the details of the spacer grids were not included, but a conservative analysis has been carried out based on the current main geometrical and physical features. Reactivity feedback, as well as axial power profile, were not included in this analysis. Results indicate that critical conditions, with clad temperatures around ∼900 °C, are reached with blockage larger than 30% in terms of area fraction. Two main effects can be distinguished: a local effect in the wake/recirculation region downstream the blockage and a global effect due to the lower mass flow rate in the blocked subchannels; the former effect gives rise to a temperature peak behind the blockage and it is dominant for large blockages (>20%), while the latter effect determines a temperature peak at the end of the active region and it is dominant for small blockages (<10%). The blockage area was placed at

Preliminary CFD analysis methodology for flow in a LFR fuel assembly

International Nuclear Information System (INIS)

Catana, A.; Ioan, M.; Serbanel, M.

2013-01-01

In this paper a preliminary Computational Fluid Dynamics (CFD) analysis was performed in order to setup a methodology to be used for more complex coolant flow analysis inside ALFRED nuclear reactor fuel assembly. The core contains 171 separated fuel assembly, each consisting in a hexagonal array of 127 fuel rods. Three honey comb spacer grids are proposed along fuel rods with the aim to keep flow geometry intact during reactor operation. The main goal of this paper is to compute some hydraulic parameters: pressure, velocity, wall shear stress and turbulence parameters with and without spacer grids. In this analysis we consider an adiabatic case, so far no heat transfer is considered but we pave the road toward more complex thermo hydraulic analysis for ALFRED (LFR in general). The CAELINUX CFD distribution was used with its main components: Salome-Meca (for geometry and mesh) and Code-Saturne as mono-phase CFD solver. Paraview and Visist Postprocessors were used for data extraction and graphical displays. (authors)

Self-assembly of coiled coil peptides into nanoparticles vs 2-d plates: effects of assembly pathway

Science.gov (United States)

Kim, Kyunghee; Pochan, Darrin

Molecular solution assembly, or self-assembly, is a process by which ordered nanostructures or patterns are formed by non-covalent interactions during assembly. Biomimicry, the use of bioinspired molecules or biologically relevant materials, is an important area of self-assembly research with peptides serving a critical role as molecular tools. The morphology of peptide assemblies can be controlled by adjusting solution conditions such as the concentration of peptides, the temperature, and pH. Herein, spherical nanostructures, which have potential for creating an encapsulation system, are formed by self-assembly when coiled coil peptides are combined in solution. These peptides are homotrimeric and heterodimeric coiled-coil bundles and the homotrimer is connected with each of heterodimer through their external surfaces via disulfide bonds. The resultant covalent constructs could co-assemble into complementary trimeric hubs, respectively. The two peptide constructs are directly mixed and assembled in solution in order to produce either spherical particles or 2-d plates depending on the solution conditions and kinetic pathway of assembly. In particular, structural changes of the self-assembled peptides are explored by control of the thermal history of the assembly solution.

Impact analysis of the spacer grid assembly and shape optimization of the attached spring

Energy Technology Data Exchange (ETDEWEB)

Park, K. J.; Lee, Z. N. [Hanyang University, Seoul (Korea)

2002-04-01

Spacer grids support fuel rods and maintain geometry from external impact loads. A simulation is performed for the strength of a spacer grid under the impact load. The critical impact load that leads to plastic deformation is identified by a free-fall test. A finite element model is established for the nonlinear simulation of the impact process. The simulation model is tuned based on the free-fall test. The model considers the aspects of welding and the contacts between components. Nonlinear finite element analysis is carried out using a software system called ABAQUS/EXPLICIT. The results are discussed from a design viewpoint. Design requirements are defined and a design process is established. The design process includes mathematical optimization as well as practical design method. The shape of the grid spring is designed to maintain its function during the lifetime of the fuel assembly. A structural optimization method is employed for the shape design. A good design is found. Commercial codes are utilized for structural analysis and optimization. 18 refs., 61 figs., 3 tabs. (Author)

Existing experimental criticality data applicable to nuclear-fuel-transportation systems

International Nuclear Information System (INIS)

Bierman, S.R.

1983-02-01

Analytical techniques are generally relied upon in making criticality evaluations involving nuclear material outside reactors. For these evaluations to be accepted the calculations must be validated by comparison with experimental data for a known set of conditions having physical and neutronic characteristics similar to those conditions being evaluated analytically. The purpose of this report is to identify those existing experimental data that are suitable for use in verifying criticality calculations on nuclear fuel transportation systems. In addition, near term needs for additional data in this area are identified. Of the considerable amount of criticality data currently existing, that are applicable to non-reactor systems, those particularly suitable for use in support of nuclear material transportation systems have been identified and catalogued into the following groups: (1) critical assemblies of fuel rods in water; (2) critical assemblies of fuel rods in water containing soluble neutron absorbers; (3) critical assemblies containing solid neutron absorber; (4) critical assemblies of fuel rods in water with heavy metal reflectors; and (5) critical assemblies of fuel rods in water with irregular features. A listing of the current near term needs for additional data in each of the groups has been developed for future use in planning criticality research in support of nuclear fuel transportation systems. The criticality experiments needed to provide these data are briefly described and identified according to priority and relative cost of performing the experiments

ZPR-3 Assembly 6F : A spherical assembly of highly enriched uranium, depleted uranium, aluminum and steel with an average {sup 235}U enrichment of 47 atom %.

Energy Technology Data Exchange (ETDEWEB)

Lell, R. M.; McKnight, R. D; Schaefer, R. W.; Nuclear Engineering Division

2010-09-30

Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 6 consisted of six phases, A through F. In each phase a critical configuration was constructed to simulate a very simple shape such as a slab, cylinder or sphere that could be analyzed with the limited analytical tools available in the 1950s. In each case the configuration consisted of a core region of metal plates surrounded by a thick depleted uranium metal reflector. The average compositions of the core configurations were essentially identical in phases A - F. ZPR-3

Engineering Mathematical Analysis Method for Productivity Rate in Linear Arrangement Serial Structure Automated Flow Assembly Line

Directory of Open Access Journals (Sweden)

Tan Chan Sin

2015-01-01

Full Text Available Productivity rate (Q or production rate is one of the important indicator criteria for industrial engineer to improve the system and finish good output in production or assembly line. Mathematical and statistical analysis method is required to be applied for productivity rate in industry visual overviews of the failure factors and further improvement within the production line especially for automated flow line since it is complicated. Mathematical model of productivity rate in linear arrangement serial structure automated flow line with different failure rate and bottleneck machining time parameters becomes the basic model for this productivity analysis. This paper presents the engineering mathematical analysis method which is applied in an automotive company which possesses automated flow assembly line in final assembly line to produce motorcycle in Malaysia. DCAS engineering and mathematical analysis method that consists of four stages known as data collection, calculation and comparison, analysis, and sustainable improvement is used to analyze productivity in automated flow assembly line based on particular mathematical model. Variety of failure rate that causes loss of productivity and bottleneck machining time is shown specifically in mathematic figure and presents the sustainable solution for productivity improvement for this final assembly automated flow line.

Self assembly of anisotropic particles with critical Casimir forces

NARCIS (Netherlands)

Nguyễn, Trúc Anh

2016-01-01

Building new materials with structures on the micron and nanoscale presents a grand challenge currently. It requires fine control in the assembly of well-designed building blocks, and understanding of the mechanical, thermodynamic, and opto-electronic properties of the resulting structures. Patchy

Teaching For Art Criticism: Incorporating Feldman’s Critical Analysis Learning Model In Students’ Studio Practice

Directory of Open Access Journals (Sweden)

Maithreyi Subramaniam

2016-01-01

Full Text Available This study adopted 30 first year graphic design students’ artwork, with critical analysis using Feldman’s model of art criticism. Data were analyzed quantitatively; descriptive statistical techniques were employed. The scores were viewed in the form of mean score and frequencies to determine students’ performances in their critical ability. Pearson Correlation Coefficient was used to find out the correlation between students’ studio practice and art critical ability scores. The findings showed most students performed slightly better than average in the critical analyses and performed best in selecting analysis among the four dimensions assessed. In the context of the students’ studio practice and critical ability, findings showed there are some connections between the students’ art critical ability and studio practice.

Potential risk of a criticality event during refuelling

Energy Technology Data Exchange (ETDEWEB)

Laurioux, V.; Deschamps, P

2003-01-01

Following a core unloading due to an unscheduled shutdown during the cycle of Dampierre unit 4, the plant operator had to operate the refuelling in its previous configuration. During this operation, the fuel assembly concerned by loading step no. 25 was left in the fuel building and the assembly for the following step, no. 26, was placed in the reactor vessel instead of the previous one, causing an irregularity in core pattern only detected at step no. 139. At that point, the refuelling machine operator realized that he was handling an assembly equipped with its rod cluster, whereas according to his handling sheet, the assembly should not be equipped with. Thanks to the favourable conditions (primary system boron concentration = 2345 ppm and cycle burn-up = 2 GWj/tU), the core remained subcritical. However, the incident analysis revealed a potential criticality risk under less favourable circumstances, i.e. refuelling with fresh 1. cycle fuel assemblies and a primary system boron concentration at the lower limit of the range allowed by the Technical Operating Specifications (2000 ppm). Moreover, further studies have shown that the two neutron source range channels (SRC) are able to provide a count rate (reflecting neutron flux) but would not detect a local increase in reactivity under refuelling conditions (normal or defective), unless a reactive pattern was formed in the immediate vicinity of one of the two channels. However, if the criticality is reached, the SRCs are able to diagnose the critical state from the significant power level of 0.1% rated power. The human error that had led to the administrative validation of a fuel handling step that had not been physically carried out highlighted both the fragility of organizational lines of defence and the inadequacy of I and C systems with regard to technical in-depth defence lines. Following these observations, and beyond the preventive measures supposed to limit the occurrence of a divergence during refuelling

Numerical Analysis on the Free Fall Motion of the Control Rod Assembly for the Sodium Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Oh, Se-Hong; Choi, Choengryul; Son, Sung-Man [ELSOLTEC, Yongin (Korea, Republic of); Kim, Jae-Yong; Yoon, Kyung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

On receiving the scram signal, the control rod assemblies are released to fall into the reactor core by its weight. Thus drop time and falling velocity of the control rod assembly must be estimated for the safety evaluation. However, because of its complex shape, it is difficult to estimate the drop time by theoretical method. In this study, numerical analysis has been carried out in order to estimate drop time and falling velocity of the control rod assembly to provide the underlying data for the design optimization. Numerical analysis has been carried out to estimate the drop time and falling velocity of the control rod assembly for sodium-cooled fast reactor. Before performing the numerical analysis for the control rod assembly, sphere dropping experiment has been carried out for verification of the CFD methodology. The result of the numerical analysis for the method verification is almost same as the result of the experiment. Falling velocity and drag force increase rapidly in the beginning. And then it goes to the stable state. When the piston head of the control rod assembly is inserted into the damper, the drag force increases instantaneously and the falling velocity decreases quickly. The falling velocity is reduced about 14 % by damper. The total drop time of the control rod assembly is about 1.47s. In the next study, the experiment for the control rod assembly will be carried out, and its result is going to be compared with the CFD analysis result.

Analysis of mixed oxide fuel critical experiments with neutronics analysis codes for boiling water reactors

International Nuclear Information System (INIS)

Tamitani, Masashi; Maruyama, Hiromi; Ishii, Kazuya; Izutsu, Sadayuki; Yamaguchi, Masao

2000-01-01

Critical experiments of UO 2 and full mixed oxide (MOX) fuel cores conducted at the Tank-type Critical Assembly (TCA) were analyzed using BWR design-purpose codes HINES and CERES with ENDF/B files and Monte Carlo fine analysis codes VMONT and MVP with the JENDL-3.2 library. The averaged values of the multiplication factors calculated with HINES/CERES, VMONT and MVP agreed with those of experiments within 0.3%Δk. The values by the design-purpose codes showed a small difference of 0.1%Δk between UO 2 and MOX cores. Monte Carlo code results showed that the JENDL-3.2 library had a tendency to overestimate the multiplication factors of UO 2 cores by about 0.3%Δk compared with those values of MOX cores. The root mean square errors of calculated power distributions were less than 1% for HINES/CERES and VMONT. These results showed that (1) the accuracy of these codes when applied to full MOX cores was almost the same as their accuracy for UO 2 cores, which confirmed the accuracy of present core design codes for full MOX cores; and (2) the accuracy of the 190-energy-group Monte Carlo calculation code VMONT was almost the same as that of the continuous-energy Monte Carlo calculation code MVP. (author)

3D Assembly Group Analysis for Cognitive Automation

Directory of Open Access Journals (Sweden)

Christian Brecher

2012-01-01

Full Text Available A concept that allows the cognitive automation of robotic assembly processes is introduced. An assembly cell comprised of two robots was designed to verify the concept. For the purpose of validation a customer-defined part group consisting of Hubelino bricks is assembled. One of the key aspects for this process is the verification of the assembly group. Hence a software component was designed that utilizes the Microsoft Kinect to perceive both depth and color data in the assembly area. This information is used to determine the current state of the assembly group and is compared to a CAD model for validation purposes. In order to efficiently resolve erroneous situations, the results are interactively accessible to a human expert. The implications for an industrial application are demonstrated by transferring the developed concepts to an assembly scenario for switch-cabinet systems.

Simulation of dielectrophoretic assembly of carbon nanotubes using 3D finite element analysis

International Nuclear Information System (INIS)

Berger, S D; McGruer, N E; Adams, G G

2015-01-01

One of the most important methods for selective and repeatable assembly of carbon nanotubes (CNTs) is alternating current dielectrophoresis (DEP). This method has been demonstrated experimentally as a viable technique for nano-scale manufacturing of novel CNT based devices. Previous numerical analyses have studied the motion of nanotubes, the volume from which they are assembled, and the rate of assembly, but have been restricted by various simplifying assumptions. In this paper we present a method for simulating the motion and behavior of CNTs subjected to dielectrophoresis using a three-dimensional electrostatic finite element analysis. By including the CNT in the finite element model, we can accurately predict the effect of the CNT on the electric field and the resulting force distribution across the CNT can be determined. We have used this information to calculate the motion of CNTs assembling onto the electrodes, and show how they tend to move towards the center of an electrode and come into contact at highly skewed angles. Our analysis suggests that the CNTs move to the electrode gap only after initially contacting the electrodes. We have also developed a model of the elastic deformation of CNTs as they approach the electrodes demonstrating how the induced forces can significantly alter the CNT shape during assembly. These results show that the CNT does not behave as a rigid body when in close proximity to the electrodes. In the future this method can be applied to a variety of real electrode geometries on a case-by-case basis and will provide more detailed insight into the specific motion and assembly parameters necessary for effective DEP assembly. (paper)

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Analysis of the rotation accident of assemblies in boiling water reactors

International Nuclear Information System (INIS)

Becerril-Gonzalez M, J. J.; Fuentes M, L.; Castillo M, J. A.; Ortiz S, J. J.; Perusquia de Cueto, R.

2012-10-01

For this work was analyzed the impact that would cause the load of a rotated fuel assembly in the behaviour of the core in the Cycle 14 of the Unit 1 of the nuclear power plant of Laguna Verde. To carry out this analysis the code Simulate-3 was used, with which was possible to analyze the behavior of the effective multiplication factor and the thermal limits (MAPRAT, MFLPD and MFLCPR). The rotation of fuel assemblies to 90, 180 and 270 grades was analyzed with regard to the design position, with 0, 1, 2 and 3 burnt cycles for these assemblies. The results show that the thermal limits remain inside the allowed values, therefore if this accident type happened the reactor could continue operating in a sure way. (Author)

Determination of fuel assembly vibrational modes through analysis of incore detector noise

International Nuclear Information System (INIS)

Johnson, R.S.

1986-01-01

In order to better characterize fuel assembly vibration at Duke Power Company's Oconee Nuclear Station, incore noise data were acquired an analyzed from prompt responding detectors incorporated in the Oconee 2, Cycle 7 core. Duke Power Company began actively pursuing an inhouse Neutron Noise Analysis program for routine surveillance of reactor internals vibration in 1979. Noise data has since been acquired and analyzed for twelve cycles of operation for the three Oconee units. Duke Power's Oconee Unit 2 is a Babcock and Wilcoxs pressurized water reactor with a rate thermal power of 2568MW. For Oconee 2, Cycle 7 operation, two test assemblies, each employing a string of seven axially-spaced, prompt responding hafnium detectors, were included in the final core design. Incore detector noise data were obtained during Cycle 7 at approximately 281 and 430 effective full power days (EFPD). In addition to the incore test detector signals, noise signals from the upper and lower chambers of the four excore power range detectors were recorded to aid in the analysis. The comparison of RMS signal levels for each incore detector and the phase relationships between detector locations within two test assemblies identified the first four fuel assembly bending modes associated with fixed end conditions

Critical seeding density improves properties and translatability of self-assembling anatomically shaped knee menisci

Science.gov (United States)

Hadidi, Pasha; Yeh, Timothy C.; Hu, Jerry C.; Athanasiou, Kyriacos A.

2014-01-01

A recent development in the field of tissue engineering is the rise of all-biologic, scaffold-free engineered tissues. Since these biomaterials rely primarily upon cells, investigation of initial seeding densities constitutes a particularly relevant aim for tissue engineers. In this study, a scaffold-free method was used to create fibrocartilage in the shape of the rabbit knee meniscus. The objectives of this study were: (i) to determine the minimum seeding density, normalized by an area of 44 mm2, necessary for the self-assembling process of fibrocartilage to occur, (ii) examine relevant biomechanical properties of engineered fibrocartilage, such as tensile and compressive stiffness and strength, and their relationship to seeding density, and (iii) identify a reduced, or optimal, number of cells needed to produce this biomaterial. It was found that a decreased initial seeding density, normalized by the area of the construct, produced superior mechanical and biochemical properties. Collagen per wet weight, glycosaminoglycans per wet weight, tensile properties, and compressive properties were all significantly greater in the 5 million cells per construct group as compared to the historical 20 million cells per construct group. Scanning electron microscopy demonstrated that a lower seeding density results in a denser tissue. Additionally, the translational potential of the self-assembling process for tissue engineering was improved though this investigation, as fewer cells may be used in the future. The results of this study underscore the potential for critical seeding densities to be investigated when researching scaffold-free engineered tissues. PMID:25234157

Sub-channel analysis of LBE-cooled fuel assemblies of accelerator driven systems

International Nuclear Information System (INIS)

Cheng, X.; Hwang, D.H.

2005-01-01

In the frame of the European PDS-XADS project, two concepts of the sub-critical reactor core cooled by liquid lead-bismuth eutectic (LBE) were proposed. In this paper, the local thermal-hydraulic behavior of both LBE-cooled fuel assemblies was analyzed. For this purpose, the sub-channel analysis code MATRA was selected, and modification was made for its applications to XADS conditions. Compared to the small core concept, the large core concept has a much lower temperatures of coolant, cladding and fuel pins. This enables a short-term realization of the core design using available technologies. The high power density of the small core results in high local temperatures of coolant, cladding and fuel. Both coolant velocity and cladding temperature are such that special attention has to be paid to avoid corrosion and erosion damage of cladding materials. A parametric study shows that under the parameters considered, mixing coefficient has the biggest effect on the coolant temperature distribution, whereas the cladding temperature is strongly affected by the selection of heat transfer correlations. (author)

Colorimetric Analysis on Flocculation of Bioinspired Au Self-Assembly for Biophotonic Application

Directory of Open Access Journals (Sweden)

Wan-Joong Kim

2009-01-01

Full Text Available Gold nanoparticles exhibited strong surface plasmon absorption and couplings between neighboring particles within bioactivated self-assembly modified their optical properties. Colorimetric analysis on the optical modification of surface plasmon resoanance (SPR shift and flocculation parameter functionalized bioinspired gold assembly for biophotonic application. The physical origin of bioinspired gold aggregation-induced shifting, decreasing, or broadening of the plasmon absorption spectra could be explained in terms of dynamic depolarization, collisional damping, and shadowing effects.

Experiments for IFR fuel criticality in ZPPR-21

International Nuclear Information System (INIS)

Olsen, D.N.; Collins, P.J.; Carpenter, S.G.

1991-01-01

A series of benchmark measurements was made in ZPPR-21 to validate criticality calculations for fuel processing operations for Argonne's Integral Fast Reactor program. Six different mixtures of Pu/U/Zr fuel with a graphite reflector were built and criticality was determined by period measurements. The assemblies were isolated from room return neutrons by a lithium hydride shield. Analysis was done using a fully-detailed model with the VIM Monte Carlo code and ENDF/B-V.2 data. Sensitivity analysis was used to validate the measurements against other benchmark data. A simple RZ model was defined and used with the KENO code. Corrections to the RZ model were provided by the VIM calculations with low statistical uncertainty. (Author)

Numerical investigations on the effect of the axial interval between intensifying spacer grids on the critical heat flux value for fuel assemblies with non-uniform axial power distribution

International Nuclear Information System (INIS)

Kireeva, D.; Oleksyuk, D.

2015-01-01

In this paper a number of numerical studies on intensifying heat exchange conducted by NRC 'Kurchatov Institute' are presented. A standardised heat exchange intensifying spacer grid (UDRI) can be installed at any height along the fuel assembly (FA) heat-generating section. When installed at the bottom of a fuel assembly, the UDRI facilitates intensive coolant mixing; the UDRI mounted at the top of a FA provides better mixing and the enhancement in heat exchange. The application of the heat exchange intensifying spacer grids results in better flattening of the coolant parameters along the cross-section and higher critical heat flux ratio. The investigations were carried out by means of numerical code SC-INT using mesh generation that have been specially designed by NRC 'Kurchatov Institute' to perform calculations for fuel assemblies equipped with the intensifying spacer grids. The effect of the axial interval between UDRI grids on the critical heat flux value for two typical axial power shapes has been investigated. The derived optimal solutions for the positioning of intensifying grids are also presented

Uncertainty analysis in Monte Carlo criticality computations

International Nuclear Information System (INIS)

Qi Ao

2011-01-01

Highlights: ► Two types of uncertainty methods for k eff Monte Carlo computations are examined. ► Sampling method has the least restrictions on perturbation but computing resources. ► Analytical method is limited to small perturbation on material properties. ► Practicality relies on efficiency, multiparameter applicability and data availability. - Abstract: Uncertainty analysis is imperative for nuclear criticality risk assessments when using Monte Carlo neutron transport methods to predict the effective neutron multiplication factor (k eff ) for fissionable material systems. For the validation of Monte Carlo codes for criticality computations against benchmark experiments, code accuracy and precision are measured by both the computational bias and uncertainty in the bias. The uncertainty in the bias accounts for known or quantified experimental, computational and model uncertainties. For the application of Monte Carlo codes for criticality analysis of fissionable material systems, an administrative margin of subcriticality must be imposed to provide additional assurance of subcriticality for any unknown or unquantified uncertainties. Because of a substantial impact of the administrative margin of subcriticality on economics and safety of nuclear fuel cycle operations, recently increasing interests in reducing the administrative margin of subcriticality make the uncertainty analysis in criticality safety computations more risk-significant. This paper provides an overview of two most popular k eff uncertainty analysis methods for Monte Carlo criticality computations: (1) sampling-based methods, and (2) analytical methods. Examples are given to demonstrate their usage in the k eff uncertainty analysis due to uncertainties in both neutronic and non-neutronic parameters of fissionable material systems.

Criticality analysis of a spent fuel shipping cask

International Nuclear Information System (INIS)

Pena, J.

1984-01-01

Criticality analysis for a system yields to the determination of the multiplication factor. Should such analysis be performed for a spent fuel shipping cask some standards must be accomplished. In this study a sample design is analyzed and criticality results are presented. (author)

Conceptual design of a digital control system for nuclear criticality experiments

International Nuclear Information System (INIS)

Rojas, S.P.

1994-04-01

Nuclear criticality is a concern in many areas of nuclear engineering including waste management, nuclear weapons testing and design, basic nuclear research, and nuclear reactor design and analysis. As in many areas of science and engineering, experimental work conducted in this field has provided a wealth of data and insight essential to the formulation of theory and the advancement in knowledge of fissioning systems. In light of the many diverse applications of nuclear criticality, there is a continuing interest to learn and understand more about the fundamental physical processes through continued experimentation. This thesis addresses the problem of setting up and programming a microprocessor-based digital control system (PLC) for a proposed critical experiment using, among other devices, a stepper motor, a joystick control mechanism, and switches. This experiment represents a revised configuration to test cylindrical nuclear waste packages. A Monte Carlo numerical study for the proposed critical assembly has been performed in order to illustrate how results from numerical calculations are used in the process of assembling the control system and to corroborate previous experimental data. In summary, a control system utilizing some common devices necessary to perform a critical experiment (stepper motor, push-buttons, etc.) has been assembled. Control components were sized using the results of a probabilistic computer code (MCNP). Finally, a program was written that illustrates the coupling between the hardware and the devices being controlled in the new test fixture

Application of international safeguards to fast critical assembly facilities. FY 1980 summary report

International Nuclear Information System (INIS)

1980-12-01

Nuclear materials inventory-verification techniques for large split-table fast critical assemblies are being studied in this program. Emphasis has been given to techniques that minimize fuel handling in order to reduce facility downtime and radiation exposure to the inventory team. The techniques studied include drawer seals, autoradiography, and spectral index measurements. Two-drawer sealing techniques have been studied, and the relative strengths and weaknesses are pointed out. The rod-type locking mechanism would not disrupt the reactor cooling air flow or interfere with autoradiography but is more expensive to implement. Passive autoradiography was used in a ZPPR inventory to verify to a 93% confidence level that less than 8-kg Pu was missing. The inventory was completed in four days by a five-member team with radiation exposures well within acceptable limits. Two autoradiographic film packages were developed to distinguish HEU from a DU matrix. The 30-mil pack requires an exposure between 4 and 16 hours and fits into most of the drawers. The 40-mil pack requires only a two-hour exposure but fits into less than half the drawers

Measurement of the ^235mU Production Cross Section Using a Critical Assembly*

Science.gov (United States)

Macri, Robert; Authier, Nicolas; Becker, John; Belier, Gilbert; Bond, Evelyn; Bredeweg, Todd; Glover, S.; Meot, Vincent; Rundberg, Robert; Vieira, David; Wilhelmy, Jerry

2006-10-01

Measurements of the creation and destruction cross sections for actinide nuclei constitute an important experimental effort in support of Stockpile Stewardship. In this talk I will give a progress report on the effort to measure the production cross section of the ^235mU isomer integrated over a fission neutron spectrum. This ongoing experiment is fielded at CEA in Valduc, France, taking advantage of the CALIBAN critical assembly. This effort is performed in collaboration with LANL, LLNL, Bruyeres le Chatel, and Valduc staff. This experiment utilizes a technique to measure internal conversion electrons from the ^235mU isomer with the French BIII detector (Bruyeres le Chatel), and involves a substantial chemistry effort (LANL) to prepare targets for irradiation and counting, as well as to remove fission fragments after irradiation. Experimental techniques will be discussed and preliminary data presented. *Work performed under the auspices of the U.S. Department of Energy by Los Alamos National Laboratory (W-7405-ENG-36) and Lawrence Livermore National Laboratory (W-7405-ENG-48), and CEA-DAM under CEA-DAM NNSA-DOE agreement.

High fidelity analysis of BWR fuel assembly with COBRA-TF/PARCS and trace codes

International Nuclear Information System (INIS)

Abarca, A.; Miro, R.; Barrachina, T.; Verdu, G.; Soler, A.

2013-01-01

The growing importance of detailed reactor core and fuel assembly description for light water reactors (LWRs) as well as the sub-channel safety analysis requires high fidelity models and coupled neutronic/thermalhydraulic codes. Hand in hand with advances in the computer technology, the nuclear safety analysis is beginning to use a more detailed thermal hydraulics and neutronics. Previously, a PWR core and a 16 by 16 fuel assembly models were developed to test and validate our COBRA-TF/PARCS v2.7 (CTF/PARCS) coupled code. In this work, a comparison of the modeling and simulation advantages and disadvantages of modern 10 by 10 BWR fuel assembly with CTF/PARCS and TRACE codes has been done. The objective of the comparison is making known the main advantages of using the sub-channel codes to perform high resolution nuclear safety analysis. The sub-channel codes, like CTF, permits obtain accurate predictions, in two flow regime, of the thermalhydraulic parameters important to safety with high local resolution. The modeled BWR fuel assembly has 91 fuel rods (81 full length and 10 partial length fuel rods) and a big square central water rod. This assembly has been modeled with high level of detail with CTF code and using the BWR modeling parameters provided by TRACE. The same neutronic PARCS's model has been used for the simulation with both codes. To compare the codes a coupled steady state has be performed. (author)

Capacity analysis of automatic transport systems in an assembly factory

NARCIS (Netherlands)

Zijm, W.H.M.; Lenstra, J.K.; Tijms, H.C.; Volgenant, A.

1989-01-01

We describe a case study concerning the capacity analysis of a completely automated transport system in a flexible assembly environment. Basically, the system is modelled as a network of queues, however, due to its complex nature, product-form network theory is not applicable. Instead, we present an

OECD/NEA burnup credit criticality benchmarks phase IIIA: Criticality calculations of BWR spent fuel assemblies in storage and transport

Energy Technology Data Exchange (ETDEWEB)

Okuno, Hiroshi; Naito, Yoshitaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ando, Yoshihira [Toshiba Corp., Kawasaki, Kanagawa (Japan)

2000-09-01

The report describes the final results of Phase IIIA Benchmarks conducted by the Burnup Credit Criticality Calculation Working Group under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA). The benchmarks are intended to confirm the predictive capability of the current computer code and data library combinations for the neutron multiplication factor (k{sub eff}) of a layer of irradiated BWR fuel assembly array model. In total 22 benchmark problems are proposed for calculations of k{sub eff}. The effects of following parameters are investigated: cooling time, inclusion/exclusion of FP nuclides and axial burnup profile, and inclusion of axial profile of void fraction or constant void fractions during burnup. Axial profiles of fractional fission rates are further requested for five cases out of the 22 problems. Twenty-one sets of results are presented, contributed by 17 institutes from 9 countries. The relative dispersion of k{sub eff} values calculated by the participants from the mean value is almost within the band of {+-}1%{delta}k/k. The deviations from the averaged calculated fission rate profiles are found to be within {+-}5% for most cases. (author)

Discovery of genes related to insecticide resistance in Bactrocera dorsalis by functional genomic analysis of a de novo assembled transcriptome.

Science.gov (United States)

Hsu, Ju-Chun; Chien, Ting-Ying; Hu, Chia-Cheng; Chen, Mei-Ju May; Wu, Wen-Jer; Feng, Hai-Tung; Haymer, David S; Chen, Chien-Yu

2012-01-01

Insecticide resistance has recently become a critical concern for control of many insect pest species. Genome sequencing and global quantization of gene expression through analysis of the transcriptome can provide useful information relevant to this challenging problem. The oriental fruit fly, Bactrocera dorsalis, is one of the world's most destructive agricultural pests, and recently it has been used as a target for studies of genetic mechanisms related to insecticide resistance. However, prior to this study, the molecular data available for this species was largely limited to genes identified through homology. To provide a broader pool of gene sequences of potential interest with regard to insecticide resistance, this study uses whole transcriptome analysis developed through de novo assembly of short reads generated by next-generation sequencing (NGS). The transcriptome of B. dorsalis was initially constructed using Illumina's Solexa sequencing technology. Qualified reads were assembled into contigs and potential splicing variants (isotigs). A total of 29,067 isotigs have putative homologues in the non-redundant (nr) protein database from NCBI, and 11,073 of these correspond to distinct D. melanogaster proteins in the RefSeq database. Approximately 5,546 isotigs contain coding sequences that are at least 80% complete and appear to represent B. dorsalis genes. We observed a strong correlation between the completeness of the assembled sequences and the expression intensity of the transcripts. The assembled sequences were also used to identify large numbers of genes potentially belonging to families related to insecticide resistance. A total of 90 P450-, 42 GST-and 37 COE-related genes, representing three major enzyme families involved in insecticide metabolism and resistance, were identified. In addition, 36 isotigs were discovered to contain target site sequences related to four classes of resistance genes. Identified sequence motifs were also analyzed to

Assembly, Annotation, and Analysis of Multiple Mycorrhizal Fungal Genomes

Energy Technology Data Exchange (ETDEWEB)

Initiative Consortium, Mycorrhizal Genomics; Kuo, Alan; Grigoriev, Igor; Kohler, Annegret; Martin, Francis

2013-03-08

Mycorrhizal fungi play critical roles in host plant health, soil community structure and chemistry, and carbon and nutrient cycling, all areas of intense interest to the US Dept. of Energy (DOE) Joint Genome Institute (JGI). To this end we are building on our earlier sequencing of the Laccaria bicolor genome by partnering with INRA-Nancy and the mycorrhizal research community in the MGI to sequence and analyze dozens of mycorrhizal genomes of all Basidiomycota and Ascomycota orders and multiple ecological types (ericoid, orchid, and ectomycorrhizal). JGI has developed and deployed high-throughput sequencing techniques, and Assembly, RNASeq, and Annotation Pipelines. In 2012 alone we sequenced, assembled, and annotated 12 draft or improved genomes of mycorrhizae, and predicted ~;;232831 genes and ~;;15011 multigene families, All of this data is publicly available on JGI MycoCosm (http://jgi.doe.gov/fungi/), which provides access to both the genome data and tools with which to analyze the data. Preliminary comparisons of the current total of 14 public mycorrhizal genomes suggest that 1) short secreted proteins potentially involved in symbiosis are more enriched in some orders than in others amongst the mycorrhizal Agaricomycetes, 2) there are wide ranges of numbers of genes involved in certain functional categories, such as signal transduction and post-translational modification, and 3) novel gene families are specific to some ecological types.

Characterization of the CALIBAN Critical Assembly Neutron Spectra using Several Adjustment Methods Based on Activation Foils Measurement

Science.gov (United States)

Casoli, Pierre; Grégoire, Gilles; Rousseau, Guillaume; Jacquet, Xavier; Authier, Nicolas

2016-02-01

CALIBAN is a metallic critical assembly managed by the Criticality, Neutron Science and Measurement Department located on the French CEA Center of Valduc. The reactor is extensively used for benchmark experiments dedicated to the evaluation of nuclear data, for electronic hardening or to study the effect of the neutrons on various materials. Therefore CALIBAN irradiation characteristics and especially its central cavity neutron spectrum have to be very accurately evaluated. In order to strengthen our knowledge of this spectrum, several adjustment methods based on activation foils measurements are being studied for a few years in the laboratory. Firstly two codes included in the UMG package have been tested and compared: MAXED and GRAVEL. More recently, the CALIBAN cavity spectrum has been studied using CALMAR, a new adjustment tool currently under development at the CEA Center of Cadarache. The article will discuss and compare the results and the quality of spectrum rebuilding obtained with the UMG codes and with the CALMAR software, from a set of activation measurements carried out in the CALIBAN irradiation cavity.

Characterization of the CALIBAN Critical Assembly Neutron Spectra using Several Adjustment Methods Based on Activation Foils Measurement

Directory of Open Access Journals (Sweden)

Casoli Pierre

2016-01-01

Full Text Available CALIBAN is a metallic critical assembly managed by the Criticality, Neutron Science and Measurement Department located on the French CEA Center of Valduc. The reactor is extensively used for benchmark experiments dedicated to the evaluation of nuclear data, for electronic hardening or to study the effect of the neutrons on various materials. Therefore CALIBAN irradiation characteristics and especially its central cavity neutron spectrum have to be very accurately evaluated. In order to strengthen our knowledge of this spectrum, several adjustment methods based on activation foils measurements are being studied for a few years in the laboratory. Firstly two codes included in the UMG package have been tested and compared: MAXED and GRAVEL. More recently, the CALIBAN cavity spectrum has been studied using CALMAR, a new adjustment tool currently under development at the CEA Center of Cadarache. The article will discuss and compare the results and the quality of spectrum rebuilding obtained with the UMG codes and with the CALMAR software, from a set of activation measurements carried out in the CALIBAN irradiation cavity.

Teaching for Art Criticism: Incorporating Feldman's Critical Analysis Learning Model in Students' Studio Practice

Science.gov (United States)

Subramaniam, Maithreyi; Hanafi, Jaffri; Putih, Abu Talib

2016-01-01

This study adopted 30 first year graphic design students' artwork, with critical analysis using Feldman's model of art criticism. Data were analyzed quantitatively; descriptive statistical techniques were employed. The scores were viewed in the form of mean score and frequencies to determine students' performances in their critical ability.…

Analysis of dismantling possibility and unloading efforts of fuel assemblies from core of WWER

International Nuclear Information System (INIS)

Danilov, V.; Dobrov, V.; Semishkin, V.; Vasilchenko, I.

2006-01-01

The computation methods of optimal dismantling sequence of fuel assemblies (FA) from core of WWER after different operating periods and accident conditions are considered. The algorithms of fuel dismantling sequence are constructed both on the basis of analysis of mutual spacer grid overlaps of adjacent fuel assemblies and numerical structure analysis of efforts required for FA removal as FA heaving from the core. Computation results for core dismantling sequence after 3-year operating period and LB LOCA are presented in the paper

The National Criticality Experiments Research Center at the Device Assembly Facility, Nevada National Security Site: Status and Capabilities, Summary Report

International Nuclear Information System (INIS)

Bragg-Sitton, S.; Bess, J.; Werner, J.

2011-01-01

The National Criticality Experiments Research Center (NCERC) was officially opened on August 29, 2011. Located within the Device Assembly Facility (DAF) at the Nevada National Security Site (NNSS), the NCERC has become a consolidation facility within the United States for critical configuration testing, particularly those involving highly enriched uranium (HEU). The DAF is a Department of Energy (DOE) owned facility that is operated by the National Nuclear Security Agency/Nevada Site Office (NNSA/NSO). User laboratories include the Lawrence Livermore National Laboratory (LLNL) and Los Alamos National Laboratory (LANL). Personnel bring their home lab qualifications and procedures with them to the DAF, such that non-site specific training need not be repeated to conduct work at DAF. The NNSS Management and Operating contractor is National Security Technologies, LLC (NSTec) and the NNSS Safeguards and Security contractor is Wackenhut Services. The complete report provides an overview and status of the available laboratories and test bays at NCERC, available test materials and test support configurations, and test requirements and limitations for performing sub-critical and critical tests. The current summary provides a brief summary of the facility status and the method by which experiments may be introduced to NCERC.

Benchmark criticality experiments for fast fission configuration with high enriched nuclear fuel

International Nuclear Information System (INIS)

Sikorin, S.N.; Mandzik, S.G.; Polazau, S.A.; Hryharovich, T.K.; Damarad, Y.V.; Palahina, Y.A.

2014-01-01

Benchmark criticality experiments of fast heterogeneous configuration with high enriched uranium (HEU) nuclear fuel were performed using the 'Giacint' critical assembly of the Joint Institute for Power and Nuclear Research - Sosny (JIPNR-Sosny) of the National Academy of Sciences of Belarus. The critical assembly core comprised fuel assemblies without a casing for the 34.8 mm wrench. Fuel assemblies contain 19 fuel rods of two types. The first type is metal uranium fuel rods with 90% enrichment by U-235; the second one is dioxide uranium fuel rods with 36% enrichment by U-235. The total fuel rods length is 620 mm, and the active fuel length is 500 mm. The outer fuel rods diameter is 7 mm, the wall is 0.2 mm thick, and the fuel material diameter is 6.4 mm. The clad material is stainless steel. The side radial reflector: the inner layer of beryllium, and the outer layer of stainless steel. The top and bottom axial reflectors are of stainless steel. The analysis of the experimental results obtained from these benchmark experiments by developing detailed calculation models and performing simulations for the different experiments is presented. The sensitivity of the obtained results for the material specifications and the modeling details were examined. The analyses used the MCNP and MCU computer programs. This paper presents the experimental and analytical results. (authors)

CFD Analysis for Advanced Integrated Head Assembly

Energy Technology Data Exchange (ETDEWEB)

Jo, Won Ho; Kang, Tae Kyo; Cho, Yeon Ho; Kim, Hyun Min [KEPCO Engineering and Construction Co., Daejeon (Korea, Republic of)

2016-10-15

The Integrated Head Assembly (IHA) is permanently installed on the reactor vessel closure head during the normal plant operation and refueling operation. It consists of a number of systems and components such as the head lifting system, seismic support system, Control Element Drive Mechanism (CEDM) cooling system, cable support system, cooling shroud assemblies. With the operating experiences of the IHA, the needs for the design change to the current APR1400 IHA arouse to improve the seismic resistance and to accommodate the convenient maintenance. In this paper, the effects of the design changes were rigorously studied for the various sizes of the inlet openings to assure the proper cooling of the CEDMs. And the system pressure differentials and required flow rate for the CEDM cooling fan were analyzed regarding the various operating conditions for determining the capacity of the fan. As a part of the design process of the AIHA, the number of air inlets and baffle regions are reduced by simplifying the design of the APR1400 IHA. The design change of the baffle regions has been made such that the maximum possible space are occupied inside the IHA cooling shroud shell while avoiding the interference with CEDMs. So, only the air inlet opening was studied for the design change to supply a sufficient cooling air flow for each CEDM. The size and location of the air inlets in middle cooling shroud assembly were determined by the CFD analyses of the AIHA. And the case CFD analyses were performed depending on the ambient air temperature and fan operating conditions. The size of the air inlet openings is increased by comparison with the initial AIHA design, and it is confirmed that the cooling air flow rate for each CEDM meet the design requirement of 800 SCFM ± 10% with the increased air inlets. At the initial analysis, the fan outlet flow rate was assumed as 48.3 lbm/s, but the result revealed that the less outflow rate at the fan is enough to meet the design requirement

Genome Assembly and Computational Analysis Pipelines for Bacterial Pathogens

KAUST Repository

Rangkuti, Farania Gama Ardhina

2011-06-01

Pathogens lie behind the deadliest pandemics in history. To date, AIDS pandemic has resulted in more than 25 million fatal cases, while tuberculosis and malaria annually claim more than 2 million lives. Comparative genomic analyses are needed to gain insights into the molecular mechanisms of pathogens, but the abundance of biological data dictates that such studies cannot be performed without the assistance of computational approaches. This explains the significant need for computational pipelines for genome assembly and analyses. The aim of this research is to develop such pipelines. This work utilizes various bioinformatics approaches to analyze the high-­throughput genomic sequence data that has been obtained from several strains of bacterial pathogens. A pipeline has been compiled for quality control for sequencing and assembly, and several protocols have been developed to detect contaminations. Visualization has been generated of genomic data in various formats, in addition to alignment, homology detection and sequence variant detection. We have also implemented a metaheuristic algorithm that significantly improves bacterial genome assemblies compared to other known methods. Experiments on Mycobacterium tuberculosis H37Rv data showed that our method resulted in improvement of N50 value of up to 9697% while consistently maintaining high accuracy, covering around 98% of the published reference genome. Other improvement efforts were also implemented, consisting of iterative local assemblies and iterative correction of contiguated bases. Our result expedites the genomic analysis of virulent genes up to single base pair resolution. It is also applicable to virtually every pathogenic microorganism, propelling further research in the control of and protection from pathogen-­associated diseases.

Thermohydraulic analysis of BWR and PWR spent fuel assemblies contained within square canisters

International Nuclear Information System (INIS)

Wiles, L.E.; McCann, R.A.

1981-09-01

This report presents the results of several thermohydraulic simulations of spent fuel assembly/canister configurations performed in support of a program investigating the feasibility of storing spent nuclear fuel assemblies in canisters that would be stored in an air environment. Eleven thermohydraulic simulations were performed. Five simulations were performed using a single BWR fuel assembly/canister design. The various cases were defined by changing the canister spacing and the heat generation rate of the fuel assembly. For each simulation a steady-state thermohydraulic solution was achieved for the region inside the canister. Similarly, six simulations were performed for a single PWR fuel assembly/canister design. The square fuel rod arrays were contained in square canisters which would permit closer packing of the canisters in a storage facility. However, closer packing of the canisters would result in higher fuel temperatures which would possibly have an adverse impact on fuel integrity. Thus, the most important aspect of the analysis was to define the peak fuel assembly temperatures for each case. These results are presented along with various temperature profiles, heat flux distributions, and air velocity profiles within the canister. 48 figures, 4 tables

Criticality impacts on LWR fuel storage efficiency

International Nuclear Information System (INIS)

Napolitano, D.

1992-01-01

This presentation discusses the criticality impacts throughout storage of fuel onsite including new fuel storage, spent fuel storage, consolidation, and dry storage. The general principles for criticality safety are also be discussed. There is first an introduction which explains today's situation for criticality safety concerns. This is followed by a discussion of criticality safety Regulatory Guides, safety limits and fundamental principles. Design objectives for criticality safety in the 1990's include higher burnups, longer cycles, and higher enrichments which impact the criticality safety design. Criticality safety for new fuel storage, spent fuel storage, fuel consolidation, and dry storage are followed by conclusions. Today's situation is one in which the US does not reprocess, and does not have an operating MRS facility or repository. High density fuel storage rack designs of the 1980s, are filling up. Dry cask storage systems for spent fuel storage are being utilized. Enrichments continue to increase PWR fuel assemblies with enrichments of 4.5 to 5.0 weight percent U-235 and BWR fuel assemblies with enrichments of 3.25 to 3.5 weight percent U-235 are common. Criticality concerns affect the capacity and the economics of light water reactor (LWR) fuel storage arrays by dictating the spacing of fuel assemblies in a storage system, or the use of poisons or exotic materials in the storage system design

Evaluation of the Cow Rumen Metagenome: Assembly by Single Copy Gene Analysis and Single Cell Genome Assemblies (Metagenomics Informatics Challenges Workshop: 10K Genomes at a Time)

Energy Technology Data Exchange (ETDEWEB)

Sczyrba, Alex

2011-10-13

DOE JGI's Alex Sczyrba on "Evaluation of the Cow Rumen Metagenome" and "Assembly by Single Copy Gene Analysis and Single Cell Genome Assemblies" at the Metagenomics Informatics Challenges Workshop held at the DOE JGI on October 12-13, 2011.

Rethinking Critical Mathematics: A Comparative Analysis of Critical, Reform, and Traditional Geometry Instructional Texts

Science.gov (United States)

Brantlinger, Andrew

2011-01-01

This paper presents findings from a comparative analysis of three similar secondary geometry texts, one critical unit, one standards-based reform unit, and one specialist chapter. I developed the critical unit as I took the tenets of critical mathematics (CM) and substantiated them in printed curricular materials in which to teach as part of a…

A method for development of efficient 3D models for neutronic calculations of ASTRA critical facility using experimental information

Energy Technology Data Exchange (ETDEWEB)

Balanin, A. L.; Boyarinov, V. F.; Glushkov, E. S.; Zimin, A. A.; Kompaniets, G. V.; Nevinitsa, V. A., E-mail: Neviniza-VA@nrcki.ru; Moroz, N. P.; Fomichenko, P. A.; Timoshinov, A. V. [National Research Center Kurchatov Institute (Russian Federation); Volkov, Yu. N. [National Research Nuclear University MEPhI (Russian Federation)

2016-12-15

The application of experimental information on measured axial distributions of fission reaction rates for development of 3D numerical models of the ASTRA critical facility taking into account azimuthal asymmetry of the assembly simulating a HTGR with annular core is substantiated. Owing to the presence of the bottom reflector and the absence of the top reflector, the application of 2D models based on experimentally determined buckling is impossible for calculation of critical assemblies of the ASTRA facility; therefore, an alternative approach based on the application of the extrapolated assembly height is proposed. This approach is exemplified by the numerical analysis of experiments on measurement of efficiency of control rods mockups and protection system (CPS).

A method for development of efficient 3D models for neutronic calculations of ASTRA critical facility using experimental information

International Nuclear Information System (INIS)

Balanin, A. L.; Boyarinov, V. F.; Glushkov, E. S.; Zimin, A. A.; Kompaniets, G. V.; Nevinitsa, V. A.; Moroz, N. P.; Fomichenko, P. A.; Timoshinov, A. V.; Volkov, Yu. N.

2016-01-01

The application of experimental information on measured axial distributions of fission reaction rates for development of 3D numerical models of the ASTRA critical facility taking into account azimuthal asymmetry of the assembly simulating a HTGR with annular core is substantiated. Owing to the presence of the bottom reflector and the absence of the top reflector, the application of 2D models based on experimentally determined buckling is impossible for calculation of critical assemblies of the ASTRA facility; therefore, an alternative approach based on the application of the extrapolated assembly height is proposed. This approach is exemplified by the numerical analysis of experiments on measurement of efficiency of control rods mockups and protection system (CPS).

Positioning of Nuclear Fuel Assemblies by Means of Image Analysis on Tomographic Data

International Nuclear Information System (INIS)

Troeng, Mats

2005-06-01

A tomographic measurement technique for nuclear fuel assemblies has been developed at the Department of Radiation Sciences at Uppsala University. The technique requires highly accurate information about the position of the measured nuclear fuel assembly relative to the measurement equipment. In experimental campaigns performed earlier, separate positioning measurements have therefore been performed in connection to the tomographic measurements. In this work, another positioning approach has been investigated, which requires only the collection of tomographic data. Here, a simplified tomographic reconstruction is performed, whereby an image is obtained. By performing image analysis on this image, the lateral and angular position of the fuel assembly can be determined. The position information can then be used to perform a more accurate tomographic reconstruction involving detailed physical modeling. Two image analysis techniques have been developed in this work. The stability of the two techniques with respect to some central parameters has been studied. The agreement between these image analysis techniques and the previously used positioning technique was found to meet the desired requirements. Furthermore, it has been shown that the image analysis techniques offer more detailed information than the previous technique. In addition, its off-line analysis properties reduce the need for valuable measurement time. When utilizing the positions obtained from the image analysis techniques in tomographic reconstructions of the rod-by-rod power distribution, the repeatability of the reconstructed values was improved. Furthermore, the reconstructions resulted in better agreement to theoretical data

Validation of the ABBN/CONSYST constants system. Part 1: Validation through the critical experiments on compact metallic cores

International Nuclear Information System (INIS)

Ivanova, T.T.; Manturov, G.N.; Nikolaev, M.N.; Rozhikhin, E.V.; Semenov, M.Yu.; Tsiboulia, A.M.

1999-01-01

Worldwide compilation of criticality safety benchmark experiments, evaluated due to an activity of the International Criticality Safety Benchmark Evaluation Project (ICSBEP), discovers new possibilities for validation of the ABBN-93.1 cross section library for criticality safety analysis. Results of calculations of small assemblies with metal-fuelled cores are presented in this paper. It is concluded that ABBN-93.1 predicts criticality of such systems with required accuracy

Experiments for IFR fuel criticality in ZPPR-21

International Nuclear Information System (INIS)

Olsen, D.N.; Collins, P.J.; Carpenter, S.G.

1991-01-01

A series of benchmark measurements was made in ZPPR-21 to validate criticality calculations for fuel operations in Argonne's Integral Fast Reactor. Six different mixtures of Pu/U/Zr fuel with a graphite reflector were built and criticality was determined by period measurements. The assemblies were isolated from room return problems by a lithium hydride shield. Analysis was done using a fully-detailed model with the VIM Monte Carlo code and ENDF/B-V.2 data. Sensitivity analysis was used to validate the measurements against other benchmark data. A simple RZ model was defined the used with the KENO code. Corrections to the RZ model were provided by the VIM calculations with low statistical uncertainty. 7 refs., 5 figs., 5 tabs

Analysis of Critical Parts and Materials

Science.gov (United States)

1980-12-01

1 1 1% 1% 1% 1% Large Orders Manual Ordering of Some Critical Parts Order Spares with Original Order Incentives Belter Capital Investment...demand 23 Large orders 24 Long lead procurement funding (including raw materials, facility funding) 25 Manpower analysis and training 26 Manual ... ordering of some critical parts 27 More active role in schedule negotiation 28 Multiple source procurements 29 Multi-year program funding 30 Order

Plutonium assemblies in reload 1 of the Dodewaard Reactor

International Nuclear Information System (INIS)

Bairiot, H.; Deramaix, P.; Vandenberg, C.; Leenders, L.; Mostert, P.

1977-01-01

Since 1963, Belgonucleaire has been developing the design of plutonium assemblies of the island type (i.e., plutonium rods inserted in the control zone of the assembly and enriched uranium rods at the periphery) for light water reactors. The application to boiling water reactors (BWRs) led to the introduction, in April 1971, of two prototype plutonium island assemblies in the Dodewaard BWR (The Netherlands): Those assemblies incorporating plutonium in 42 percent of the rods are interchangeable with standard uranium assemblies of the same reload. Their design, which had to meet these criteria, was performed using the routine order in use at Belgonucleaire; experimental checks included a mock-up configuration simulated in the VENUS critical facility at Mol and open-vessel cold critical experiments performed in the Dodewaard core. The pelleted plutonium rods were fabricated and controlled by Belgonucleaire following the manufacturing procedures developed at the production plant. In one of the assemblies, three vibrated plutonium fuel rods with a lower fuel density were introduced in the three most highly rated positions to reduce the power rating. Those plutonium assemblies experienced peak pellet ratings up to 535 W/cm and were discharged in April 1974 after having reached a mean burnup of approximately 21,000 MWd/MT. In-core instrumentation during operation, visual examinations, and reactivity substitution experiments during reactor shutdown did not indicate any special feature for those assemblies compared to the standard uranium assemblies, thereby demonstrating their interchangeability

Impact of axial burnup profile on criticality safety of ANPP spent fuel cask

International Nuclear Information System (INIS)

Bznuni, S.

2006-01-01

Criticality safety assessment for WWER-440 NUHOMS cask with spent nuclear fuel from Armenian NPP has been performed. The cask was designed in such way that the neutron multiplication factor k eff must be below 0,95 for all operational modes and accident conditions. Usually for criticality analysis, fresh fuel approach with the highest enrichment is taken as conservative assumption as it was done for ANPP. NRSC ANRA in order to improve future fuel storage efficiency initiated research with taking into account burn up credit in the criticality safety assessment. Axial burn up profile (end effect) has essential impact on criticality safety justification analysis. However this phenomenon was not taken into account in the Safety Analysis Report of NUHOMS spent fuel storage constructed on the site of ANPP. Although ANRA does not yet accept burn up credit approach for ANPP spent fuel storage, assessment of impact of axial burnup profile on criticality of spent fuel assemblies has important value for future activities of ANRA. This paper presents results of criticality calculations of spent fuel assemblies with axial burn up profile. Horizontal burn up profile isn't taken account since influence of the horizontal variation of the burn up is much less than the axial variation. The actinides and actinides + fission products approach are discussed. The calculations were carried out with STARBUCS module of SCALE 5.0 code package developed at Oak Ridge National laboratory. SCALE5.0 sequence CSAS26 (KENO-VI) was used for evaluation the k eff for 3-D problems. Obtained results showed that criticality of ANPP spent fuel cask is very sensitive to the end effect

Burnup credit calculations for criticality safety justification for RBMK-1000 spent fuel of transport and storage systems

Directory of Open Access Journals (Sweden)

V. V. Galchenko

2010-12-01

Full Text Available In present paper the burnup credit calculations for TK-8 transport container and SVJP-1 spent fuel storage fa-cility of pool type with RBMK-1000 spent fuel during 100-years of cooling time were performed for criticality safety analysis purpose using MCNP and SCALE codes. Only actinides were taken into account for these critical systems. Two approaches were analyzed with isotopes distribution calculations along fuel assembly height and without it. The results show that subcriticality margin is increased considerably using burnup credit and isotopes distribution along fuel assembly height made this value more reasonable.

Material selection and assembly method of battery pack for compact electric vehicle

Science.gov (United States)

Lewchalermwong, N.; Masomtob, M.; Lailuck, V.; Charoenphonphanich, C.

2018-01-01

Battery packs become the key component in electric vehicles (EVs). The main costs of which are battery cells and assembling processes. The battery cell is indeed priced from battery manufacturers while the assembling cost is dependent on battery pack designs. Battery pack designers need overall cost as cheap as possible, but it still requires high performance and more safety. Material selection and assembly method as well as component design are very important to determine the cost-effectiveness of battery modules and battery packs. Therefore, this work presents Decision Matrix, which can aid in the decision-making process of component materials and assembly methods for a battery module design and a battery pack design. The aim of this study is to take the advantage of incorporating Architecture Analysis method into decision matrix methods by capturing best practices for conducting design architecture analysis in full account of key design components critical to ensure efficient and effective development of the designs. The methodology also considers the impacts of choice-alternatives along multiple dimensions. Various alternatives for materials and assembly techniques of battery pack are evaluated, and some sample costs are presented. Due to many components in the battery pack, only seven components which are positive busbar and Z busbar are represented in this paper for using decision matrix methods.

The Influencing Factor Analysis on the Performance Evaluation of Assembly Line Balancing Problem Level 1 (SALBP-1) Based on ANOVA Method

Science.gov (United States)

Chen, Jie; Hu, Jiangnan

2017-06-01

Industry 4.0 and lean production has become the focus of manufacturing. A current issue is to analyse the performance of the assembly line balancing. This study focus on distinguishing the factors influencing the assembly line balancing. The one-way ANOVA method is applied to explore the significant degree of distinguished factors. And regression model is built to find key points. The maximal task time (tmax ), the quantity of tasks (n), and degree of convergence of precedence graph (conv) are critical for the performance of assembly line balancing. The conclusion will do a favor to the lean production in the manufacturing.

Sub-channel analysis of a HPLWR fuel assembly with STAR-CD

International Nuclear Information System (INIS)

Himmel, Steffen R.; Class, Andreas G.; Schulenberg, Thomas; Laurien, Eckart

2008-01-01

Hofmeister et. al. developed a first design proposal for a HPLWR fuel assembly, consisting of a square 7 by 7 fuel pin arrangement within an assembly box and a water box in the centre, replacing 9 fuel rods. Instead of conventional grid spacers, wire wraps are considered due to good coolant mixing and low pressure drop in either flow direction. Within the present work, a novel approach describing the coolant heat up in the sub-channels of such an assembly has been investigated: the commercial software package STAR-CD has been used as a sub-channel code to investigate the thermal-hydraulic performance of such an HPLWR fuel assembly. The aim of the work is to demonstrate that a widely accepted commercial Computational Fluid Dynamics (CFD) code can be used for full rod bundle analysis by applying minor modifications to it. In steady of writing a dedicated code system with numerical solver routines and post-processing tools for sub-channel analyses, the user benefits from the optimized Graphical User Interface (GUI) already provided in STAR-CD. Moreover, a smooth transition to full three-dimensional modeling of the fluid flow inside rod bundles will be possible with the same code system, if considered to be necessary, just by refining the spatial discretization. Steady-state and transient flow regimes can be studied for design as well as reactor safety analysis. As the STAR-CD code uses the Finite Volume Method (FVM) for spatial discretization, the conservation equations for mass, momentum and energy were modified via user-subroutines to obtain the equations known from the usual sub-channel approach. The method will be explained in detail and results will be discussed. (author)

Equation of State for Phospholipid Self-Assembly

DEFF Research Database (Denmark)

Marsh, Derek

2016-01-01

Phospholipid self-assembly is the basis of biomembrane stability. The entropy of transfer from water to self-assembled micelles of lysophosphatidylcholines and diacyl phosphatidylcholines with different chain lengths converges to a common value at a temperature of 44°C. The corresponding enthalpies...... of transfer converge at ∼-18°C. An equation of state for the free energy of self-assembly formulated from this thermodynamic data depends on the heat capacity of transfer as the sole parameter needed to specify a particular lipid. For lipids lacking calorimetric data, measurement of the critical micelle...

Technical specifications for the Oak Ridge Critical Experiments Facility

International Nuclear Information System (INIS)

Stinnett, R.M.

1986-01-01

These Technical Specifications for the Oak Ridge Critical Experiments Facility (CEF) delineate limiting conditions of operation for the facility. The CEF is used primarily for testing the High Flux Isotope Reactor (HFIR) fuel assemblies. Specifically, the Criticality Testing Unit, Liquid (CTUL), located in the CEF, is used for the HFIR fuel assembly test. The test is performed to satisfy the surveillance requirements of the HFIR Technical Specifications. The test is used to determine the water-submerged shutdown margin for each fuel assembly. 11 refs

Does College Teach Critical Thinking? A Meta-Analysis

Science.gov (United States)

Huber, Christopher R.; Kuncel, Nathan R.

2016-01-01

Educators view critical thinking as an essential skill, yet it remains unclear how effectively it is being taught in college. This meta-analysis synthesizes research on gains in critical thinking skills and attitudinal dispositions over various time frames in college. The results suggest that both critical thinking skills and dispositions improve…

Row of fuel assemblies analysis under seismic loading: Modelling and experimental validation

International Nuclear Information System (INIS)

Ricciardi, Guillaume; Bellizzi, Sergio; Collard, Bruno; Cochelin, Bruno

2009-01-01

The aim of this study was to develop a numerical model for predicting the impact behaviour at fuel assembly level of a whole reactor core under seismic loading conditions. This model was based on a porous medium approach accounting for the dynamics of both the fluid and structure, which interact. The fluid is studied in the whole reactor core domain and each fuel assembly is modelled in the form of a deformable porous medium with a nonlinear constitutive law. The contact between fuel assemblies is modelled in the form of elastic stops, so that the impact forces can be assessed. Simulations were performed to predict the dynamics of a six fuel assemblies row immersed in stagnant water and the whole apparatus was placed on a shaking table mimicking seismic loading conditions. The maximum values of the impact forces predicted by the model were in good agreement with the experimental data. A Proper Orthogonal Decomposition analysis was performed on the numerical data to analyse the mechanical behaviour of the fluid and structure more closely.

Multi-objective Analysis for a Sequencing Planning of Mixed-model Assembly Line

Science.gov (United States)

Shimizu, Yoshiaki; Waki, Toshiya; Yoo, Jae Kyu

Diversified customer demands are raising importance of just-in-time and agile manufacturing much more than before. Accordingly, introduction of mixed-model assembly lines becomes popular to realize the small-lot-multi-kinds production. Since it produces various kinds on the same assembly line, a rational management is of special importance. With this point of view, this study focuses on a sequencing problem of mixed-model assembly line including a paint line as its preceding process. By taking into account the paint line together, reducing work-in-process (WIP) inventory between these heterogeneous lines becomes a major concern of the sequencing problem besides improving production efficiency. Finally, we have formulated the sequencing problem as a bi-objective optimization problem to prevent various line stoppages, and to reduce the volume of WIP inventory simultaneously. Then we have proposed a practical method for the multi-objective analysis. For this purpose, we applied the weighting method to derive the Pareto front. Actually, the resulting problem is solved by a meta-heuristic method like SA (Simulated Annealing). Through numerical experiments, we verified the validity of the proposed approach, and discussed the significance of trade-off analysis between the conflicting objectives.

SCALE 5: Powerful new criticality safety analysis tools

International Nuclear Information System (INIS)

Bowman, Stephen M.; Hollenbach, Daniel F.; Dehart, Mark D.; Rearden, Bradley T.; Gauld, Ian C.; Goluoglu, Sedat

2003-01-01

Version 5 of the SCALE computer software system developed at Oak Ridge National Laboratory, scheduled for release in December 2003, contains several significant new modules and sequences for criticality safety analysis and marks the most important update to SCALE in more than a decade. This paper highlights the capabilities of these new modules and sequences, including continuous energy flux spectra for processing multigroup problem-dependent cross sections; one- and three-dimensional sensitivity and uncertainty analyses for criticality safety evaluations; two-dimensional flexible mesh discrete ordinates code; automated burnup-credit analysis sequence; and one-dimensional material distribution optimization for criticality safety. (author)

CFD Analysis of Hot Spot Fuel Temperature in the Control Fuel Block Assembly of a VHTR core

International Nuclear Information System (INIS)

Kim, Min Hwan; Tak, Nam Il; Noh, Jae Man

2010-01-01

The Very High Temperature Reactor (VHTR) dedicated for efficient hydrogen production requires core outlet temperatures of more than 950 .deg. C. As the outlet temperature increases, the thermal margin of the core decreases, which highlights the need for a detailed analysis to reduce its uncertainty. Tak et al. performed CFD analysis for a 1/12 fuel assembly model and compared the result with a simple unit-cell model in order to emphasize the need of a detailed CFD analysis for the prediction of hot spot fuel temperatures. Their CFD model, however, was focused on the standard fuel assembly but not on the control fuel assembly in which a considerable amount of bypass flow is expected to occur through the control rod passages. In this study, a CFD model for the control fuel block assembly is developed and applied for the hot spot analyses of PMR200 core. Not only the bypass flow but also the cross flow is considered in the analyses

A modification design and adjusting test for instruments and control system of critical assembly

International Nuclear Information System (INIS)

Wu Manrong; Li Guangjian

1996-12-01

A more reliable and safe control system and it's instruments for HFETRCA (high flux engineering test reactor critical assembly) have been built. In the system high performance CMOS unit was used, which has high integration, strong anti-interference and high trigger threshold. In the design of control rod driving circuit, the speed negative feedback principle was applied that results in more stable rotating rate of motors of transmission mechanism and more flexibility of adjusting rod speed. In order to improve reactor safety in accident, additional control circuit is equipped, by which not only control rods with electromagnet will rapidly drop but also other control rods will insert at the speed of 2∼6 times faster than the normal inserting speed. The key technique in the adjustment and new method of anti-interference are also introduced. After more than 40 times physical experiments with (4 x 4 - 4) fuel element in HFETRC, it is proved that the design and adjustment of the system is successful and they can be used as a reference to others. (3 figs., 2 tabs.)

Computational methods for nuclear criticality safety analysis

International Nuclear Information System (INIS)

Maragni, M.G.

1992-01-01

Nuclear criticality safety analyses require the utilization of methods which have been tested and verified against benchmarks results. In this work, criticality calculations based on the KENO-IV and MCNP codes are studied aiming the qualification of these methods at the IPEN-CNEN/SP and COPESP. The utilization of variance reduction techniques is important to reduce the computer execution time, and several of them are analysed. As practical example of the above methods, a criticality safety analysis for the storage tubes for irradiated fuel elements from the IEA-R1 research has been carried out. This analysis showed that the MCNP code is more adequate for problems with complex geometries, and the KENO-IV code shows conservative results when it is not used the generalized geometry option. (author)

An analysis of fast critical experiments using JEF-1-based 50-group constant set

Energy Technology Data Exchange (ETDEWEB)

Kim, Jung Do; Gil, Choong Sup; Kim, Young Cheol [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1993-09-01

JEF-1-based 50-group cross section set for fast reactor calculations was generated using NJOY system. The set was then examined by analyzing measured integral quantities such as criticality and central reaction rate ratios for 27 fast critical assemblies. The calculated results using the new set were also compared with those of ENDF/B-IV or -V based fast set. In general, the JEF-1-based set shows an improvement in predicting measured integral quantities in comparison with the previous set. With a few exceptions, JEF-1 results are compareble to those of ENDF/B-V.

Teaching For Art Criticism: Incorporating Feldman’s Critical Analysis Learning Model In Students’ Studio Practice

OpenAIRE

Maithreyi Subramaniam; Jaffri Hanafi; Abu Talib Putih

2016-01-01

This study adopted 30 first year graphic design students’ artwork, with critical analysis using Feldman’s model of art criticism. Data were analyzed quantitatively; descriptive statistical techniques were employed. The scores were viewed in the form of mean score and frequencies to determine students’ performances in their critical ability. Pearson Correlation Coefficient was used to find out the correlation between students’ studio practice and art critical ability scores. The...

Analysis of reconfigurable assembly system framing systems in automotive industry

Directory of Open Access Journals (Sweden)

Md Zain Mohamad Zamri

2017-01-01

Full Text Available Current trend in automotive industry shows increasing demand for multiple models with lean production. Prior to that, automotive manufacturing systems evolved from mass production to flexible automation. Material handling systems and equipment in a single assembly line with multiple models require high investment but with low throughput thus making production cost relatively high. Current assembly process of side structure and undercarriage with downtime occurrence during assembly process affecting production performance (quality, cost and delivery. Manufacturing facilities should allow more flexibility and increase intelligence evolving toward novel reconfigurable assembly systems (RAS. RAS is envisaged capable of increasing factor flexibility and responsiveness by incorporating assembly jig, robot and framing, which could be next generation of world class automotive assembly systems. This project research proposes a new methodology of framework reconfigurable assembly systems principles in automotive framing systems i.e. enhance assembly process between side structure assembly and undercarriage assembly which a new RAS is capable to reconfigure the assembly processes of multiple model on a single assembly line. Simulation software (Witness will be used to simulate and validate current and proposed assembly process. RAS is expected to be a solution for rapid change in structure and for a responsively adjustable production capacity. Quality, cost and delivery are production key parameters that can be achieved by implementing RAS.

Systemic analysis of the caulking assembly process

Directory of Open Access Journals (Sweden)

Rodean Claudiu

2017-01-01

Full Text Available The present paper highlights the importance of a caulking process which is nowadays less studied in comparison with the growing of its usage in the automotive industry. Due to the fact that the caulking operation is used in domains with high importance such as shock absorbers and brake systems there comes the demand of this paper to detail the parameters which characterize the process, viewed as input data and output data, and the requirements asked for the final product. The paper presents the actual measurement methods used for analysis the performance of the caulking assembly. All this parameters leads to an analysis algorithm of performance established for the caulking process which it is used later in the paper for an experimental research. The study is a basis from which it will be able to go to further researches in order to optimize the following processing.

Nondeterministic self-assembly of two tile types on a lattice.

Science.gov (United States)

Tesoro, S; Ahnert, S E

2016-04-01

Self-assembly is ubiquitous in nature, particularly in biology, where it underlies the formation of protein quaternary structure and protein aggregation. Quaternary structure assembles deterministically and performs a wide range of important functions in the cell, whereas protein aggregation is the hallmark of a number of diseases and represents a nondeterministic self-assembly process. Here we build on previous work on a lattice model of deterministic self-assembly to investigate nondeterministic self-assembly of single lattice tiles and mixtures of two tiles at varying relative concentrations. Despite limiting the simplicity of the model to two interface types, which results in 13 topologically distinct single tiles and 106 topologically distinct sets of two tiles, we observe a wide variety of concentration-dependent behaviors. Several two-tile sets display critical behaviors in the form of a sharp transition from bound to unbound structures as the relative concentration of one tile to another increases. Other sets exhibit gradual monotonic changes in structural density, or nonmonotonic changes, while again others show no concentration dependence at all. We catalog this extensive range of behaviors and present a model that provides a reasonably good estimate of the critical concentrations for a subset of the critical transitions. In addition, we show that the structures resulting from these tile sets are fractal, with one of two different fractal dimensions.

Development of whole core thermal-hydraulic analysis program ACT. 4. Simplified fuel assembly model and parallelization by MPI

International Nuclear Information System (INIS)

Ohshima, Hiroyuki

2001-10-01

A whole core thermal-hydraulic analysis program ACT is being developed for the purpose of evaluating detailed in-core thermal hydraulic phenomena of fast reactors including the effect of the flow between wrapper-tube walls (inter-wrapper flow) under various reactor operation conditions. As appropriate boundary conditions in addition to a detailed modeling of the core are essential for accurate simulations of in-core thermal hydraulics, ACT consists of not only fuel assembly and inter-wrapper flow analysis modules but also a heat transport system analysis module that gives response of the plant dynamics to the core model. This report describes incorporation of a simplified model to the fuel assembly analysis module and program parallelization by a message passing method toward large-scale simulations. ACT has a fuel assembly analysis module which can simulate a whole fuel pin bundle in each fuel assembly of the core and, however, it may take much CPU time for a large-scale core simulation. Therefore, a simplified fuel assembly model that is thermal-hydraulically equivalent to the detailed one has been incorporated in order to save the simulation time and resources. This simplified model is applied to several parts of fuel assemblies in a core where the detailed simulation results are not required. With regard to the program parallelization, the calculation load and the data flow of ACT were analyzed and the optimum parallelization has been done including the improvement of the numerical simulation algorithm of ACT. Message Passing Interface (MPI) is applied to data communication between processes and synchronization in parallel calculations. Parallelized ACT was verified through a comparison simulation with the original one. In addition to the above works, input manuals of the core analysis module and the heat transport system analysis module have been prepared. (author)

Experimental study of neutron noise with criticality safety applications in mind

International Nuclear Information System (INIS)

Barnett, C.S.

1985-11-01

A study has been conducted on the statistics of detected neutrons that leaked from four subcritical reflected, enriched-uranium assemblies, to explore the feasibility of developing a criticality warning system based on neutron noise analysis. Studies were conducted on three possible discriminators, i.e., three signatures that might be used to discriminate among assemblies of various multiplications. The noise analysis techniques studied performed well enough in deeply subcritical situations to deserve testing in an applications environment. They have a good chance of detecting changes in reactivity that are potentially dangerous. One can expect sharpest results when doing comparisons, i.e., when comparing two records, one taken in the past under circumstances known to be normal and one taken now to search for change

Optimizing Transcriptome Assemblies for Eleusine indica Leaf and Seedling by Combining Multiple Assemblies from Three De Novo Assemblers

Directory of Open Access Journals (Sweden)

Shu Chen

2015-03-01

Full Text Available Due to rapid advances in sequencing technology, increasing amounts of genomic and transcriptomic data are available for plant species, presenting enormous challenges for biocomputing analysis. A crucial first step for a successful transcriptomics-based study is the building of a high-quality assembly. Here, we utilized three different de novo assemblers (Trinity, Velvet, and CLC and the EvidentialGene pipeline tr2aacds to assemble two optimized transcript sets for the notorious weed species, . Two RNA sequencing (RNA-seq datasets from leaf and aboveground seedlings were processed using three assemblers, which resulted in 20 assemblies for each dataset. The contig numbers and N50 values of each assembly were compared to study the effect of read number, k-mer size, and in silico normalization on assembly output. The 20 assemblies were then processed through the tr2aacds pipeline to remove redundant transcripts and to select the transcript set with the best coding potential. Each assembly contributed a considerable proportion to the final transcript combination with the exception of the CLC-k14. Thus each assembler and parameter set did assemble better contigs for certain transcripts. The redundancy, total contig number, N50, fully assembled contig number, and transcripts related to target-site herbicide resistance were evaluated for the EvidentialGene and Trinity assemblies. Comparing the EvidentialGene set with the Trinity assembly revealed improved quality and reduced redundancy in both leaf and seedling EvidentialGene sets. The optimized transcriptome references will be useful for studying herbicide resistance in and the evolutionary process in the three allotetraploid offspring.

In-depth analysis of accidental criticality in a reprocessing plant

International Nuclear Information System (INIS)

Thomas, W.; Gmal, B.

1989-01-01

An in-depth-analysis including probabilistic considerations has been performed for a potential criticality excursion in two large vessels in a planned reprocessing plant. Criticality safety of these components is based on limitation of uranium and plutonium concentration. The main intention of this study was to investigate the potential and probable magnitude of a criticality in greater detail and to detect possible weaknesses of criticality safety provisions. The results of the analysis show, that the calculated probability of less than 5 · 10 -4 /a a for a criticality excursion in the most relevant rework tank may be further reduced. The peak power rework tank may be further reduced by design modifications. The peak power and total fissions of the critical excursion as assessed in a preceding analysis for licensing are conservative by factors of 10 and 4, respectively

Preliminary evaluation of pin power distribution for fuel assemblies of SMART by MCNP

International Nuclear Information System (INIS)

Kim, Kyo Youn

1998-08-01

Monte Carlo transport code MCNP can describe an object sophisticately by use of three-dimensional modelling and can adopt a continuous energy cross-section library. Therefore MCNP has been widely utilized in the field of radiation physics to estimate fluxes and dose rates for nuclear facilities and to review results from conventional methods such a as discrete ordinates method and point kernel method. The Monte Carlo method has recently been introduced to estimated the neutron multiplication factor and pin power distribution in the fuel assembly of a reactor core. The operating thermal power of SMART core is 330 MWt and there are 57 fuel assemblies in the core. In this study it was assumed that the core has 4 types of fuel assemblies. In this study, MCNP4a was used to perform to estimate criticality and normalized pin power distribution in a fuel assembly of SMART core. The results from MCNP4a calculations are able to be used review those from nuclear design/analysis code. It is very complicated to pick up interested data from MCNP output list and to normalize pin power distribution in a fuel assembly because MCNP is not only a nuclear design/analysis code. In this study a program FAPIN was developed to generated a generate a normalized pin power distribution from the MCNP output list. (author). 11 refs

Assessment of criticality safety

International Nuclear Information System (INIS)

Lloyd, R.C.; Heaberlin, S.W.; Clayton, E.D.; Carter, R.D.

1979-01-01

A study was made of 100 violations of criticality safety specifications reported over a 10-y period in the operations of fuel reprocessing plants. The seriousness of each rule violation was evaluated by assigning it a severity index value. The underlying causes or reasons, for the violations were identified. A criticality event tree was constructed using the parameters, causes, and reasons found in the analysis of the infractions. The event tree provides a means for visualizing the paths to an accidental criticality. Some 65% of the violations were caused by misinterpretation on the part of the operator, being attributed to a lack of clarity in the specification and insufficient training; 33% were attributed to lack of care, whereas only 2% were caused by mechanical failure. A fault tree was constructed by assembling the events that could contribute to an accident. With suitable data on the probabilities of contributing events, the probability of the accident's occurrence can be forecast. Estimated probabilities for criticality were made, based on the limited data available, that in this case indicate a minimum time span of 244 y of plant operation per accident ranging up to approx. 3000 y subject to the various underlying assumptions made. Some general suggestions for improvement are formulated based on the cases studied. Although conclusions for other plants may differ in detail, the general method of analysis and the fault tree logic should prove applicable. 4 figures, 8 tables

PRECLOSURE CRITICALITY ANALYSIS PROCESS REPORT

International Nuclear Information System (INIS)

Danise, A.E.

2004-01-01

This report describes a process for performing preclosure criticality analyses for a repository at Yucca Mountain, Nevada. These analyses will be performed from the time of receipt of fissile material until permanent closure of the repository (preclosure period). The process describes how criticality safety analyses will be performed for various configurations of waste in or out of waste packages that could occur during preclosure as a result of normal operations or event sequences. The criticality safety analysis considers those event sequences resulting in unanticipated moderation, loss of neutron absorber, geometric changes, or administrative errors in waste form placement (loading) of the waste package. The report proposes a criticality analyses process for preclosure to allow a consistent transition from preclosure to postclosure, thereby possibly reducing potential cost increases and delays in licensing of Yucca Mountain. The proposed approach provides the advantage of using a parallel regulatory framework for evaluation of preclosure and postclosure performance and is consistent with the U.S. Nuclear Regulatory Commission's approach of supporting risk-informed, performance-based regulation for fuel cycle facilities, ''Yucca Mountain Review Plan, Final Report'', and 10 CFR Part 63. The criticality-related criteria for ensuring subcriticality are also described as well as which guidance documents will be utilized. Preclosure operations and facilities have significant similarities to existing facilities and operations currently regulated by the U.S. Nuclear Regulatory Commission; therefore, the design approach for preclosure criticality safety will be dictated by existing regulatory requirements while using a risk-informed approach with burnup credit for in-package operations

Utilization of the MCNP-3A code for criticality safety analysis

International Nuclear Information System (INIS)

Maragni, M.G.; Moreira, J.M.L.

1996-01-01

In the last decade, Brazil started to operate facilities for processing and storing uranium in different forms. The necessity of criticality safety analysis appeared in the design phase of the uranium pilot process plants and also in the licensing of transportation and storage of fissile materials. The 2-MW research reactor and the Angra I power plant also required criticality safety assessments because their spent-fuel storage was approaching full-capacity utilization. The criticality safety analysis in Brazil has been based on KENO IV code calculations, which present some difficulties for correct geometry representation. The MCNP-3A code is not reported to be used frequently for criticality safety analysis in Brazil, but its good geometry representation makes it a possible tool for treating problems of complex geometry. A set of benchmark tests was performed to verify its applicability for criticality safety analysis in Brazil. This paper presents several benchmark tests aimed at selecting a set of options available in the MCNP-3A code that would be adequate for criticality safety analysis. The MCNP-3A code is also compared with the KENO-IV code regarding its performance for criticality safety analysis

Empowerment in critical care - a concept analysis.

Science.gov (United States)

Wåhlin, Ingrid

2017-03-01

The purpose of this paper was to analyse how the concept of empowerment is defined in the scientific literature in relation to critical care. As empowerment is a mutual process affecting all individuals involved, the perspectives of not only patients and next of kin but also staff were sought. A literature review and a concept analysis based on Walker and Avant's analysis procedure were used to identify the basic elements of empowerment in critical care. Twenty-two articles with a focus on critical care were discovered and included in the investigation. A mutual and supportive relationship, knowledge, skills, power within oneself and self-determination were found to be the common attributes of empowerment in critical care. The results could be adapted and used for all parties involved in critical care - whether patients, next of kin or staff - as these defining attributes are assumed to be universal to all three groups, even if the more specific content of each attribute varies between groups and individuals. Even if empowerment is only sparsely used in relation to critical care, it appears to be a very useful concept in this context. The benefits of improving empowerment are extensive: decreased levels of distress and strain, increased sense of coherence and control over situation, and personal and/or professional development and growth, together with increased comfort and inner satisfaction. © 2016 The Authors. Scandinavian Journal of Caring Sciences published by John Wiley & Sons Ltd on behalf of Nordic College.

Analysis of a spacecraft instrument ball bearing assembly lubricated by a perfluoroalkylether

Science.gov (United States)

Morales, W.; Jones, W. R., Jr.; Buckley, D. H.

1986-01-01

An analysis of a spacecraft instrument ball bearing assembly, subjected to a scanning life test, was performed to determine the possible case of rotational problems involving these units aboard several satellites. The analysis indicated an ineffective transfer of a fluorinated liquid lubricant from a phenolic retainer to the bearing balls. Part of the analysis led to a novel HPLC separation method employing a fluorinated mobile phase in conjunction with silica based size exclusion columns.

Reference calculations on critical assemblies with Apollo2 code working with a fine multigroup mesh

International Nuclear Information System (INIS)

Aggery, A.

1999-12-01

The objective of this thesis is to add to the multigroup transport code APOLLO2 the capability to perform deterministic reference calculations, for any type of reactor, using a very fine energy mesh of several thousand groups. This new reference tool allows us to validate the self-shielding model used in industrial applications, to perform depletion calculations, differential effects calculations, critical buckling calculations or to evaluate precisely data required by the self shielding model. At its origin, APOLLO2 was designed to perform routine calculations with energy meshes around one hundred groups. That is why, in the current format of cross sections libraries, almost each value of the multigroup energy transfer matrix is stored. As this format is not convenient for a high number of groups (concerning memory size), we had to search out a new format for removal matrices and consequently to modify the code. In the new format we found, only some values of removal matrices are kept (these values depend on a reconstruction precision choice), the other ones being reconstructed by a linear interpolation, what reduces the size of these matrices. Then we had to show that APOLLO2 working with a fine multigroup mesh had the capability to perform reference calculations on any assembly geometry. For that, we successfully carried out the validation with several calculations for which we compared APOLLO2 results (obtained with the universal mesh of 11276 groups) to results obtained with Monte Carlo codes (MCNP, TRIPOLI4). Physical analysis led with this new tool have been very fruitful and show a great potential for such an R and D tool. (author)

Self assembly of rectangular shapes on concentration programming and probabilistic tile assembly models.

Science.gov (United States)

Kundeti, Vamsi; Rajasekaran, Sanguthevar

2012-06-01

Efficient tile sets for self assembling rectilinear shapes is of critical importance in algorithmic self assembly. A lower bound on the tile complexity of any deterministic self assembly system for an n × n square is [Formula: see text] (inferred from the Kolmogrov complexity). Deterministic self assembly systems with an optimal tile complexity have been designed for squares and related shapes in the past. However designing [Formula: see text] unique tiles specific to a shape is still an intensive task in the laboratory. On the other hand copies of a tile can be made rapidly using PCR (polymerase chain reaction) experiments. This led to the study of self assembly on tile concentration programming models. We present two major results in this paper on the concentration programming model. First we show how to self assemble rectangles with a fixed aspect ratio ( α:β ), with high probability, using Θ( α + β ) tiles. This result is much stronger than the existing results by Kao et al. (Randomized self-assembly for approximate shapes, LNCS, vol 5125. Springer, Heidelberg, 2008) and Doty (Randomized self-assembly for exact shapes. In: proceedings of the 50th annual IEEE symposium on foundations of computer science (FOCS), IEEE, Atlanta. pp 85-94, 2009)-which can only self assembly squares and rely on tiles which perform binary arithmetic. On the other hand, our result is based on a technique called staircase sampling . This technique eliminates the need for sub-tiles which perform binary arithmetic, reduces the constant in the asymptotic bound, and eliminates the need for approximate frames (Kao et al. Randomized self-assembly for approximate shapes, LNCS, vol 5125. Springer, Heidelberg, 2008). Our second result applies staircase sampling on the equimolar concentration programming model (The tile complexity of linear assemblies. In: proceedings of the 36th international colloquium automata, languages and programming: Part I on ICALP '09, Springer-Verlag, pp 235

Structural Aspects of Bacterial Outer Membrane Protein Assembly.

Science.gov (United States)

Calmettes, Charles; Judd, Andrew; Moraes, Trevor F

2015-01-01

The outer membrane of Gram-negative bacteria is predominantly populated by β-Barrel proteins and lipid anchored proteins that serve a variety of biological functions. The proper folding and assembly of these proteins is essential for bacterial viability and often plays a critical role in virulence and pathogenesis. The β-barrel assembly machinery (Bam) complex is responsible for the proper assembly of β-barrels into the outer membrane of Gram-negative bacteria, whereas the localization of lipoproteins (Lol) system is required for proper targeting of lipoproteins to the outer membrane.

Critical experiments AT Pacific Northwest Laboratory

International Nuclear Information System (INIS)

Clayton, E.D.; Bierman, S.R.

1984-01-01

After a short description of the facility, a brief listing of the principal types of fuel forms and assembly geometries is provided. A number of experiments have recently been performed on plain fissionable units, or isolated assemblies of single units, that include measurements on solutions composed of Pu-U mixtures and critical experiment data on lattices of low enriched uranium in water. Experiments have been performed on planar arrays of containers with Pu solutions because of the lack of data in this field concerning the safe storage of nuclear fuel; others have been conducted on arrays of low enriched U lattice assemblies. Neutronic measurements to date have shown they can be used to provide additional benchmark data for improvement and validation of criticality codes. Studies have previously been made to ascertain the need for critical experiments in support of fuel recycle operations. The result of an effort to update the list of needed critical experiments is summarized in this section. Experiments are listed in support of uranium based fuels and fast breeder reactor fuels. An effort is made to identify those areas within the fuel cycle wherein the critical experiment data would be applied and to identify the experiments (and data) required to fulfill the needs in each of these areas. The type and form of fuel on which the data would be obtained also are identified. In presenting this information, no attempt is made to describe the experiments in detail, or to define the actual number of critical experiments that might be needed to provide the required data

Strength evaluation of top nozzle holddown spring screw for nuclear fuel assembly

International Nuclear Information System (INIS)

Koh, S. K.; Won, S. Y.; Ryu, C. H.; Kim, Y. J.; Lee, K. S.; Jeon, K. L.

2002-01-01

Holddown springs are required to maintain the nuclear fuel assembly in contact with lower core plate and permit thermal and irradiation-induced length changes. Therefore, the holddown spring screw must be designed such that it is capable of sustaining the loads imposed by the initial tensile preload and operational loads. Prior to assessing the structural integrity of the spring screw in the corrosive and irradiating environment throughout the design lifetime of the fuel assembly, the strength evaluation of screw was made in this paper using the mechanics of materials and finite element methods. Calculations based on the mechanics of materials, showed that the preloaded screw with an operating holddown force had a quite large margin of safety in strength. However, the elastic-plastic finite element analysis showed that the local stresses at the critical regions of head-shank fillet and thread root significantly exceeded than the yield strength of the screw material, resulting in local plastic deformation. Preloading on the screw applied for tightening had beneficial effects on the screw strength by reducing the stress level at the critical regions, compared to the screw without preload. Calculated spring deflection using the finite element analysis was in close agreement with the experimentally measured deflection

Static stress analysis of coupling superconducting solenoid coil assembly for muon ionization cooling experiment

International Nuclear Information System (INIS)

Pan Heng; Wang Li; Wu Hong; Guo Xinglong; Xu Fengyu

2010-01-01

The stresses in the coupling superconducting solenoid coil assembly, which is applied in the Muon Ionization Cooling Experiment (MICE), are critical for the structure design and mechanical stability because of a large diameter and relative high magnetic field. This paper presents an analytical stress solution for the MICE coupling coil assembly. The stress due to winding tension is calculated by assuming the coil package as a set of combined cylinders. The thermal and electromechanical stresses are obtained by solving the partial differential equations of displacement based on the power series expansion method. The analytical stress solution is proved to be feasible by calculating stresses in a tested superconducting solenoid with 2.58 m bore at room temperature. The analytical result of the MICE coupling coil is in good agreement with that of the finite element which shows that the transverse shear stress induced by Lorentz force is principally dominant to magnet instability. (authors)

Charged triblock copolymer self-assembly into charged micelles

Science.gov (United States)

Chen, Yingchao; Zhang, Ke; Zhu, Jiahua; Wooley, Karen; Pochan, Darrin; Department of Material Science; Engineering University of Delaware Team; Department of Chemistry Texas A&M University Collaboration

2011-03-01

Micelles were formed through the self-assembly of amphiphlic block copolymer poly(acrylic acid)-block-poly(methyl acrylate)-block-polystyrene (PAA-PMA-PS). ~Importantly, the polymer is complexed with diamine molecules in pure THF solution prior to water titration solvent processing-a critical aspect in the control of final micelle geometry. The addition of diamine triggers acid-base complexation ~between the carboxylic acid PAA side chains and amines. ~Remarkably uniform spheres were found to form close-packed patterns when forced into dried films and thin, solvated films when an excess of amine was used in the polymer assembly process. Surface properties and structural features of these hexagonal-packed spherical micelles with charged corona have been explored by various characterization methods including Transmission Electron Microscopy (TEM), cryogenic TEM, z-potential analysis and Dynamic Light Scattering. The forming mechanism for this pattern and morphology changes against external stimulate such as salt will be discussed.

Comparative analysis of methods for the microcircuit assembly on flexible polyimide carriers

Directory of Open Access Journals (Sweden)

Verbitskiy V. G.

2013-10-01

Full Text Available The article presents a classification of methods for the microcircuit assembly with the use of flexible polyimide carriers of different types, and their comparative analysis. The most appropriate method for the manufacturing of flexible dual-layer carriers is singled out.

Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Ohshima, Hiroyuki; Uwaba, Tomoyuki [Japan Atomic Energy Agency (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan); Hashimoto, Akihiko; Imai, Yasutomo [NDD Corporation (1-1-6 Jounan, Mito, Ibaraki 310-0803, Japan) (Japan); Ito, Masahiro [NESI Inc. (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan)

2015-12-31

A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

Validation of the criticality calculation for fuel elements using the Gamtec 2 - Keno 2 and 4

International Nuclear Information System (INIS)

Teixeira, M.C.C.; Andrade, M.C. de

1990-01-01

For criticality safety in the fabrication, storage and transportation of fuel assemblies, subcriticality analysis must be done. The calculations are performed at CDTN with the GAMTEC computer code, to homogenize the fuel assembly in order to create 16 group cross-section library, and with KENO code, for determining the multiplication factor. To validate the calculational method, suitable Benchmark experiments have been done. The results show that the calculational model overestimates kef when kef+ 2 σ was considered. (author) [pt

Assembly of the epithelial Na+ channel evaluated using sucrose gradient sedimentation analysis.

Science.gov (United States)

Cheng, C; Prince, L S; Snyder, P M; Welsh, M J

1998-08-28

Three subunits, alpha, beta, and gamma, contribute to the formation of the epithelial Na+ channel. To investigate the oligomeric assembly of the channel complex, we used sucrose gradient sedimentation analysis to determine the sedimentation properties of individual subunits and heteromultimers comprised of multiple subunits. When the alpha subunit was expressed alone, it first formed an oligomeric complex with a sedimentation coefficient of 11 S, and then generated a higher order multimer of 25 S. In contrast, individual beta and gamma subunits predominately assembled into 11 S complexes. We obtained similar results with expression in cells and in vitro. When we co-expressed beta with alpha or with alpha plus gamma, the beta subunit assembled into a 25 S complex. Glycosylation of the alpha subunit was not required for assembly into a 25 S complex. We found that the alpha subunit formed intra-chain disulfide bonds. Although such bonds were not required to generate an oligomeric complex, under nonreducing conditions the alpha subunit formed a complex that migrated more homogeneously at 25 S. This suggests that intra-chain disulfide bonds may stabilize the complex. These data suggest that the epithelial Na+ channel subunits form high order oligomeric complexes and that the alpha subunit contains the information that facilitates such formation. Interestingly, the ability of the alpha, but not the beta or gamma, subunit to assemble into a 25 S homomeric complex correlates with the ability of these subunits to generate functional channels when expressed alone.

The upgrade of integrity analysis module and the mechanical behavior evaluation for the assembly duct

Energy Technology Data Exchange (ETDEWEB)

Lee, Byoung Oon; Lee, Dong Uk; Kim, Young Il [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-05-01

The high neutron fluxes and operating temperatures associated with KALIMER are inducing the important radiation damage phenomena, which can cause significant dimensional changes in the core components of the reactor.The thermo-mechanical analysis of the assembly ducts for KALIMER are mainly performed to evaluate the following items.1) change of reactivity. 2) force at pads on core assemblies. 3) withdrawal force at refueling. 4) loading and refueling deviation of assembly ducts. 5) bowing modes for control assembly. In this report, the model for the evaluation of reactivity change as well as the refueling model and the withdrawl force model are upgraded. And the reactivity change is considered as the most important parameter among the above items. Therefore, the sensitivity analyses mainly associated with reactivity change are carried out. As the results, the pad gap between the assembly ducts preliminary driven for keeping the (-) reactivity change. 9 refs., 24 figs., 2 tabs. (Author)

Criticality assessment of LLRWDF closure

International Nuclear Information System (INIS)

Sarrack, A.G.; Weber, J.H.; Woody, N.D.

1992-01-01

During the operation of the Low Level Radioactive Waste Disposal Facility (LLRWDF), large amounts (greater than 100 kg) of enriched uranium (EU) were buried. This EU came primarily from the closing and decontamination of the Naval Fuels Facility in the time period from 1987--1989. Waste Management Operations (WMO) procedures were used to keep the EU boxes separated to prevent possible criticality during normal operation. Closure of the LLRWDF is currently being planned, and waste stabilization by Dynamic Compaction (DC) is proposed. Dynamic compaction will crush the containers in the LLRWDF and result in changes in their geometry. Research of the LLRWDF operations and record keeping practices have shown that the EU contents of trenches are known, but details of the arrangement of the contents cannot be proven. Reviews of the trench contents, combined with analysis of potential critical configurations, revealed that some portions of the LLRWDF can be expected to be free of criticality concerns while other sections have credible probabilities for the assembly of a critical mass, even in the uncompacted configuration. This will have an impact on the closure options and which trenches can be compacted

Critical Discourse Analysis and Leadership

Science.gov (United States)

Arriaza, Gilberto

2015-01-01

This article outlines the need of infusing critical discourse analysis into the preparation and support of prospective school leaders. It argues that in the process of school transformation, the school leader must possess the ability to self-reflect on his/her language and understand the potential power of language as a means that may support or…

Fundamentals of critical analysis: the concept of validity and analysis essentials

Directory of Open Access Journals (Sweden)

Miguel Araujo Alonso

2012-01-01

Full Text Available Critical analysis of literature is an assessment process that allows the reader to get an idea of potential error in the results of a study, errors arising either from bias or confusion. Critical analysis attempts to establish whether the study meets expected criteria or methodological conditions. There are many checklists available that are commonly used to guide this analysis, but filling out a checklist is not tantamount to critical appraisal. Internal validity is defined as the extent to which a research finding actually represents the true relationship between exposure and outcome, considering the unique conditions in which the study was carried out. Attention must be given to the inclusion and exclusion criteria that were used, on the sampling methods, on the baseline characteristics of the patients that were enrolled in the study. External validity refers to the possibility of generalizing conclusions beyond the study sample or the study population. External validity includes population validity and ecological validity. Lastly, the article covers potential threats to external validity that must be considered when analyzing a study.

30 CFR 7.45 - Critical characteristics

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Critical characteristics 7.45 Section 7.45 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR TESTING, EVALUATION, AND... characteristics The following critical characteristics shall be inspected or tested on each battery assembly to...

30 CFR 7.305 - Critical characteristics.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Critical characteristics. 7.305 Section 7.305 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR TESTING, EVALUATION, AND... characteristics. The following critical characteristics shall be inspected on each motor assembly to which an...

Criticality safety analysis of Hanford Waste Tank 241-101-SY

International Nuclear Information System (INIS)

Perry, R.T.; Sapir, J.L.; Krohn, B.J.

1993-01-01

As part of a safety assessment for proposed pump mixing operations to mitigate episodic gas releases in Tank 241-101-SY at the Hanford Site, Richland, Washington, a criticality safety analysis was made using the Sn transport code ONEDANT. The tank contains approximately one million gallons of waste and an estimated 910 G of plutonium. the criticality analysis considers reconfiguration and underestimation of plutonium content. The results indicate that Tank SY-101 does not present a criticality hazard. These methods are also used in criticality analyses of other Hanford tanks

Some neutronics of innovative subcritical assembly with fast neutron spectrum

International Nuclear Information System (INIS)

Kiyavitskaya, H.; Fokov, Yu.; Rutkovskaya, Ch.; Sadovich, S.; Kasuk, D.; Gohar, Y.; Bolshinsky, I.

2013-01-01

Conclusion: • New assembly can be used to: • develop the experimental techniques and adapt the existing ones for monitoring the sub-criticality level, neutron spectra measurements, etc; • study the spatial kinetics of sub-critical and critical systems with fast neutron spectra; • measure the transmutation reaction rates of minor-actinides etc

Calculation of the fissile mass of a graphite moderated critical assembly using 93% enriched uranium

International Nuclear Information System (INIS)

Correa, F.; Marzo, M.A.S.; Collussi, I.; Ferreira, A.C.A.

1976-01-01

The critical mass of uranium has been calculated for a graphite moderated set fueled with 93% enriched uranium to be mounted on the Instituto de Energia Atomica split table Zero Power Reactor. The core composition was optimized to permit the maximum number of configurations to be studied. Analysis of three core compositions shows that 8 Kg of uranium enriched to 93% - U-235 (by weight) and 100 Kg of thorium would be sufficient for criticality experiments [pt

Impedance Analysis of the Conditioning of PBI–Based Electrode Membrane Assemblies for High Temperature PEM Fuel Cells

DEFF Research Database (Denmark)

Araya, Samuel Simon; Vang, Jakob Rabjerg; Andreasen, Søren Juhl

2013-01-01

This work analyses the conditioning of single fuel cell assemblies based on different membrane electrode assembly (MEA) types, produced by different methods. The analysis was done by means of electrochemical impedance spectroscopy, and the changes in the fitted resistances of the all the tested...

Fuel assembly design for APR1400 with low CBC

Energy Technology Data Exchange (ETDEWEB)

Hah, Chang Joo, E-mail: changhah@kings.ac.kr [Department of NPP Engineering, KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2015-04-29

APR 1400 is a PWR (Pressurized Water Reactor) with rated power of 3983 MWth and 241 assemblies. Recently, demand for extremely longer cycle up to 24 months is increasing with challenge of higher critical boron concentration (CBC). In this paper, assembly design method of selecting Gd-rods is introduced to reduce CBC. The purpose of the method is to lower the critical boron concentration of the preliminary core loading pattern (PLP), and consequently to achieve more negative or less positive moderator temperature coefficient (MTC). In this method, both the ratio of the number of low-Gd rod to the number of high-Gd rod (r) and assembly average Gd wt% (w) are the decision variables. The target function is the amount of soluble boron concentration reduction, which can be converted to Δk{sub TARGET}. A set of new designed fuel assembly satisfies an objective function, min [f=∑{sub i}(Δk{sub FA}−Δk{sub i})], and enables a final loading pattern to reach a target CBC. The constraints required to determine a set of Δk are physically realizable pair, (r,w), and the sum of Δk of new designed assemblies as close to Δk{sub TARGET} as possible. New Gd-bearing assemblies selected based on valid pairs of (r,w) are replaced with existing assemblies in a PLP. This design methodology is applied to Shin-Kori Unit 3 Cycle 1 used as a reference model. CASMO-3/MASTER code is used for depletion calculation. CASMO-3/MASTER calculations with new designed assemblies produce lower CBC than the expected CBC, proving that the proposed method works successful.

De novo Assembly and Analysis of the Chilean Pencil Catfish Trichomycterus areolatus Transcriptome

Science.gov (United States)

Schulze, Thomas T.; Ali, Jonathan M.; Bartlett, Maggie L.; McFarland, Madalyn M.; Clement, Emalie J.; Won, Harim I.; Sanford, Austin G.; Monzingo, Elyssa B.; Martens, Matthew C.; Hemsley, Ryan M.; Kumar, Sidharta; Gouin, Nicolas; Kolok, Alan S.; Davis, Paul H.

2016-01-01

Trichomycterus areolatus is an endemic species of pencil catfish that inhabits the riffles and rapids of many freshwater ecosystems of Chile. Despite its unique adaptation to Chile's high gradient watersheds and therefore potential application in the investigation of ecosystem integrity and environmental contamination, relatively little is known regarding the molecular biology of this environmental sentinel. Here, we detail the assembly of the Trichomycterus areolatus transcriptome, a molecular resource for the study of this organism and its molecular response to the environment. RNA-Seq reads were obtained by next-generation sequencing with an Illumina® platform and processed using PRINSEQ. The transcriptome assembly was performed using TRINITY assembler. Transcriptome validation was performed by functional characterization with KOG, KEGG, and GO analyses. Additionally, differential expression analysis highlights sex-specific expression patterns, and a list of endocrine and oxidative stress related transcripts are included. PMID:27672404

Fuel Assembly Damping Summary

Energy Technology Data Exchange (ETDEWEB)

Lee, Kanghee; Kang, Heungseok; Oh, Dongseok; Yoon, Kyungho; Kim, Hyungkyu; Kim, Jaeyong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

This paper summary the fuel assembly damping data in air/in still water/under flow, released from foreign fuel vendors, compared our data with the published data. Some technical issues in fuel assembly damping measurement testing are also briefly discussed. Understanding of each fuel assembly damping mechanisms according to the surrounding medium and flow velocity can support the fuel design improvement in fuel assembly dynamics and structural integrity aspect. Because the upgraded requirements of the newly-developed advanced reactor system will demands to minimize fuel design margin in integrity evaluation, reduction in conservatism of fuel assembly damping can contribute to alleviate the fuel design margin for sure. Damping is an energy dissipation mechanism in a vibrating mechanical structure and prevents a resonant structure from having infinite vibration amplitudes. The sources of fuel assembly damping are various from support friction to flow contribution, and it can be increased by the viscosity or drag of surrounding fluid medium or the average velocity of water flowing. Fuel licensing requires fuel design evaluation in transient or accidental condition. Dynamic response analysis of fuel assembly is to show fuel integrity and requires information on assembly-wise damping in dry condition and under wet or water flowing condition. However, damping measurement test for the full-scale fuel assembly prototype is not easy to carry out because of the scale (fuel prototype, test facility), unsteadiness of test data (scattering, random sampling and processing), instrumentation under water flowing (water-proof response measurement), and noise. LWR fuel technology division in KAERI is preparing the infra structure for damping measurement test of full-scale fuel assembly, to support fuel industries and related research activities. Here is a preliminary summary of fuel assembly damping, published in the literature. Some technical issues in fuel assembly damping

Stress analysis of fuel assemblies under seismic load

International Nuclear Information System (INIS)

Kiselev, A.; Krutko, E.; Kiselev, I.; Tutnov, A.

2011-01-01

One of the important parts of fuel assemblies (FA) safety validation is their strength estimation under the dynamic loads, such as the vibration effects caused by the work of reactor units and the seismic exposure of an earthquake, leading to extreme inertia loads on all elements of the NPP. Taking into account structural features of FA and a very large mass, the exposure of seismic loads can lead to significant deformation of fuel assemblies. It is necessary to assess the magnitude of the force interaction between the FA in case of an earthquake to estimate the strength and performance of fuel assemblies. It is also necessary to compute FA bending forms and maximum values for further RPS control rods inserting time estimation, and for disassembly possibility justification of the core and individual FA after the earthquake. The problem of WWER-1000 core dynamic behavior modeling with TVS-2M fuel assemblies under the seismic loads exposure using the finite element method is described. Each fuel assembly is represented by equivalent rod finite element model. The reactor core is simulated by 163 fuel assemblies in accordance with the reactor core construction. Stiffness characteristics of fuel assemblies are determined on the results of a series of static and dynamic TVS-2M FA field tests. The special algorithm was developed to consider the fuel rod slippage effect during deformation. The special contact elements are introduced into the model of the core to take into account the interaction of fuel assemblies with their neighbors and with core barrel. Solution of the dynamic equilibrium equations system of finite element model is implemented by direct integration using the explicit scheme. Parallel algorithms for numerical integration on multiprocessor computers with graphics processing unit is developed to improve the efficiency of calculations. Values of nodes displacement in finite element model of reactor core as a function of seismic excitation time are obtained

Critical Analysis of a Website: A Critique based on Critical Applied Linguistics and Critical Discourse Analysis

Directory of Open Access Journals (Sweden)

Rina Agustina

2013-05-01

Full Text Available E-learning was easily found through browsing internet, which was mostly free of charge and provided various learning materials. Spellingcity.com was one of e-learning websites for teaching and learning English to learn spelling, vocabulary and writing, which offered various games and activities for young learners, 6 until 8 year old learners in particular. Having considered those constraints, this paper aimed to analyse the website from two different views: (1 critical applied linguistics  (CAL aspects and (2 critical  discourse analysis (CDA. After analysing the website using CAL and CDA, it was found that the website was adequate for beginner, in which it provided fun learning through games as well as challenged learners’ to test their vocabulary. Despite of these strengths, there were several issues required further thinking in terms of learners’ broad knowledge, such as, some of learning materials focused on states in America. It was quite difficult for EFL learners if they did not have adequate general knowledge. Thus, the findings implied that the website could be used as a supporting learning material, which accompanied textbooks and vocabulary exercise books.

Use of an oscillation technique to measure effective cross-sections of fissionable samples in critical assemblies

International Nuclear Information System (INIS)

Tretiakoff, O.; Vidal, R.; Carre, J.C.; Robin, M.

1964-01-01

The authors describe the technique used to measure the effective absorption and neutron-yield cross-sections of a fissionable sample. These two values are determined by analysing the signals due to the variation in reactivity (over-all signal) and the local perturbation in the flux (local signal) produced by the oscillating sample. These signals are standardized by means of a set of samples containing quantities of fissionable material ( 235 U) and an absorber, boron, which are well known. The measurements are made for different neutron spectra characterized by lattice parameters which constitute the central zone within which the sample moves. This technique is used to study the effective cross-sections of uranium-plutonium alloys for different heavy-water and graphite lattices in the MINERVE and MARIUS critical assemblies. The same experiments are carried out on fuel samples of different irradiations in order to determine the evolution of effective cross-sections as a function of the spectrum and the irradiations. (authors) [fr

Fast critical assembly safeguards: NDA methods for highly enriched uranium. Summary report, October 1978-September 1979

International Nuclear Information System (INIS)

Bellinger, F.O.; Winslow, G.H.

1980-12-01

Nondestructive assay (NDA) methods, principally passive gamma measurements and active neutron interrogation, have been studied for their safeguards effectiveness and programmatic impact as tools for making inventories of highly enriched uranium fast critical assembly fuel plates. It was concluded that no NDA method is the sole answer to the safeguards problem, that each of those emphasized here has its place in an integrated safeguards system, and that each has minimum facility impact. It was found that the 185-keV area, as determined with a NaI detector, was independent of highly-enriched uranium (HEU) plate irradiation history, though the random neutron driver methods used here did not permit accurate assay of irradiated plates. Containment procedures most effective for accurate assaying were considered, and a particular geometry is recommended for active interrogation by a random driver. A model, pertinent to that geometry, which relates the effects of multiplication and self-absorption, is described. Probabilities of failing to detect that plates are missing are examined

Measurements in Los Alamos benchmark criticals and the central reactivity discrepancy

International Nuclear Information System (INIS)

Davey, W.G.; Hansen, G.E.; Koelling, J.J.; McLaughlin, T.P.

1978-01-01

Measurements in seven Los Alamos fast critical facilities are described; all are related to elucidating the causes of the central reactivity discrepancy in fast reactors. Specific capabilities of these specialized assemblies permit measurements well-above delayed critical and these confirm the validity of the delayed neutron data used for calibration; there is therefore no reactivity-scale error. Reactivity measurements in these homogeneous assemblies exhibit no discrepancy. It is concluded that nuclear data should not be adjusted to eliminate the discrepancy found in other, heterogeneous assemblies

New self-assembly strategies for next generation lithography

Science.gov (United States)

Schwartz, Evan L.; Bosworth, Joan K.; Paik, Marvin Y.; Ober, Christopher K.

2010-04-01

Future demands of the semiconductor industry call for robust patterning strategies for critical dimensions below twenty nanometers. The self assembly of block copolymers stands out as a promising, potentially lower cost alternative to other technologies such as e-beam or nanoimprint lithography. One approach is to use block copolymers that can be lithographically patterned by incorporating a negative-tone photoresist as the majority (matrix) phase of the block copolymer, paired with photoacid generator and a crosslinker moiety. In this system, poly(α-methylstyrene-block-hydroxystyrene)(PαMS-b-PHOST), the block copolymer is spin-coated as a thin film, processed to a desired microdomain orientation with long-range order, and then photopatterned. Therefore, selfassembly of the block copolymer only occurs in select areas due to the crosslinking of the matrix phase, and the minority phase polymer can be removed to produce a nanoporous template. Using bulk TEM analysis, we demonstrate how the critical dimension of this block copolymer is shown to scale with polymer molecular weight using a simple power law relation. Enthalpic interactions such as hydrogen bonding are used to blend inorganic additives in order to enhance the etch resistance of the PHOST block. We demonstrate how lithographically patternable block copolymers might fit in to future processing strategies to produce etch-resistant self-assembled features at length scales impossible with conventional lithography.

Development of the criticality accident analysis code, AGNES

International Nuclear Information System (INIS)

Nakajima, Ken

1989-01-01

In the design works for the facilities which handle nuclear fuel, the evaluation of criticality accidents cannot be avoided even if their possibility is as small as negligible. In particular in the system using solution fuel like uranyl nitrate, solution has the property easily becoming dangerous form, and all the past criticality accidents occurred in the case of solution, therefore, the evaluation of criticality accidents becomes the most important item of safety analysis. When a criticality accident occurred in a solution fuel system, due to the generation and movement of radiolysis gas voids, the oscillation of power output and pressure pulses are observed. In order to evaluate the effect of criticality accidents, these output oscillation and pressure pulses must be calculated accurately. For this purpose, the development of the dynamic characteristic code AGNES (Accidentally Generated Nuclear Excursion Simulation code) was carried out. The AGNES is the reactor dynamic characteristic code having two independent void models. Modified energy model and pressure model, and as the benchmark calculation of the AGNES code, the results of the experimental analysis on the CRAC experiment are reported. (K.I.)

Risk and Interdependencies in Critical Infrastructures A Guideline for Analysis

CERN Document Server

Utne, Ingrid; Vatn, Jørn

2012-01-01

Today’s society is completely dependent on critical networks such as  water supply, sewage, electricity, ICT and transportation. Risk and vulnerability analyses are needed to grasp the impact of threats and hazards. However, these become quite complex as there are strong interdependencies both within and between infrastructure systems. Risk and Interdependencies in Critical Infrastructures: A  guideline for analysis provides methods for analyzing risks and interdependencies of critical infrastructures.  A number of analysis approaches are described and are adapted to each of these infrastructures. Various approaches are also revised, and all are supported by several examples and illustrations. Particular emphasis is given to the analysis of various interdependencies that often exist between the infrastructures.  Risk and Interdependencies in Critical Infrastructures: A  guideline for analysis provides a good tool to identify the hazards that are threatening your infrastructures, and will enhance the un...

OECD/NEA burnup credit criticality benchmarks phase IIIB: Burnup calculations of BWR fuel assemblies for storage and transport

International Nuclear Information System (INIS)

Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya

2002-02-01

The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD). The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated PWR fuel assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform throughout the channel box and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated atomic number densities of 12 actinides and 20 fission product nuclides were found to be for the most part within a range of ±10% relative to the average, although some results, esp. 155 Eu and gadolinium isotopes, exceeded the band, which will require further investigation. Pin-wise burnup results agreed well among the participants. The results in the infinite neutron multiplication factor k ∞ also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the averaged value noticeably for the void fraction of 70%. (author)

OECD/NEA burnup credit criticality benchmarks phase IIIB. Burnup calculations of BWR fuel assemblies for storage and transport

Energy Technology Data Exchange (ETDEWEB)

Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-02-01

The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD). The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated PWR fuel assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform throughout the channel box and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated atomic number densities of 12 actinides and 20 fission product nuclides were found to be for the most part within a range of {+-}10% relative to the average, although some results, esp. {sup 155}Eu and gadolinium isotopes, exceeded the band, which will require further investigation. Pin-wise burnup results agreed well among the participants. The results in the infinite neutron multiplication factor k{sub {infinity}} also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the averaged value noticeably for the void fraction of 70%. (author)

Numerical modeling assistance system in finite element analysis for the structure of an assembly

International Nuclear Information System (INIS)

Nakajima, Norihiro; Nishida, Akemi; Kawakami, Yoshiaki; Suzuki, Yoshio; Sawa, Kazuhiro; Iigaki, Kazuhiko

2015-01-01

The objective of structural analysis and seismic response analysis are well recognized and utilized as one of sophisticated analysis tools for design objects in the nuclear engineering. The way to design nuclear facilities is always in compromising with many index, such as costs, performance, robustness and so on, but the most important issues is the safety. It is true the structural analysis and seismic response analysis plays an important role to insure the safety, since it is well known that Japan is always facing to the earthquake. In this paper, a numerical analysis's controlling and managing system is implemented on a supercomputer, which controls the modelling process and data treating for structural robustness, although a numerical analysis's manager only controls a structural analysis by finite element method. The modeling process is described by the list of function ID and its procedures in a data base. The analytical modeling manager executes the process by order of the lists for simulation procedures. The manager controls the intention of an analysis by changing the analytical process one to another. Modeling process was experimentally found that may subject to the intention of designing index. The numerical experiments were carried out with static analyses and dynamic analyses. The results of its experiment introduce reasonable discussion to understand the accuracy of simulation. In the numerical experiments, K, supercomputer is utilized by using parallel computing resource with the controlling and managing system. The structural analysis and seismic response analysis are done by the FIEST, finite element analysis for the structure of an assembly, which carries out the simulation by gathering components. As components are attached to one another to build an assembly, and, therefore, the interactions between the components due to differences in material properties and their connection conditions considerably affect the behavior of an assembly. FIESTA is

Screening Analysis of Criticality Features, Events, and Processes for License Application

International Nuclear Information System (INIS)

J.A. McClure

2004-01-01

This report documents the screening analysis of postclosure criticality features, events, and processes. It addresses the probability of criticality events resulting from degradation processes as well as disruptive events (i.e., seismic, rock fall, and igneous). Probability evaluations are performed utilizing the configuration generator described in ''Configuration Generator Model'', a component of the methodology from ''Disposal Criticality Analysis Methodology Topical Report''. The total probability per package of criticality is compared against the regulatory probability criterion for inclusion of events established in 10 CFR 63.114(d) (consider only events that have at least one chance in 10,000 of occurring over 10,000 years). The total probability of criticality accounts for the evaluation of identified potential critical configurations of all baselined commercial and U.S. Department of Energy spent nuclear fuel waste form and waste package combinations, both internal and external to the waste packages. This criticality screening analysis utilizes available information for the 21-Pressurized Water Reactor Absorber Plate, 12-Pressurized Water Reactor Absorber Plate, 44-Boiling Water Reactor Absorber Plate, 24-Boiling Water Reactor Absorber Plate, and the 5-Defense High-Level Radioactive Waste/U.S. Department of Energy Short waste package types. Where defensible, assumptions have been made for the evaluation of the following waste package types in order to perform a complete criticality screening analysis: 21-Pressurized Water Reactor Control Rod, 5-Defense High-Level Radioactive Waste/U.S. Department of Energy Long, and 2-Multi-Canister Overpack/2-Defense High-Level Radioactive Waste package types. The inputs used to establish probabilities for this analysis report are based on information and data generated for the Total System Performance Assessment for the License Application, where available. This analysis report determines whether criticality is to be

Electron microscopic analysis of rotavirus assembly-replication intermediates

International Nuclear Information System (INIS)

Boudreaux, Crystal E.; Kelly, Deborah F.; McDonald, Sarah M.

2015-01-01

Rotaviruses (RVs) replicate their segmented, double-stranded RNA genomes in tandem with early virion assembly. In this study, we sought to gain insight into the ultrastructure of RV assembly-replication intermediates (RIs) using transmission electron microscopy (EM). Specifically, we examined a replicase-competent, subcellular fraction that contains all known RV RIs. Three never-before-seen complexes were visualized in this fraction. Using in vitro reconstitution, we showed that ~15-nm doughnut-shaped proteins in strings were nonstructural protein 2 (NSP2) bound to viral RNA transcripts. Moreover, using immunoaffinity-capture EM, we revealed that ~20-nm pebble-shaped complexes contain the viral RNA polymerase (VP1) and RNA capping enzyme (VP3). Finally, using a gel purification method, we demonstrated that ~30–70-nm electron-dense, particle-shaped complexes represent replicase-competent core RIs, containing VP1, VP3, and NSP2 as well as capsid proteins VP2 and VP6. The results of this study raise new questions about the interactions among viral proteins and RNA during the concerted assembly–replicase process. - Highlights: • Rotaviruses replicate their genomes in tandem with early virion assembly. • Little is known about rotavirus assembly-replication intermediates. • Assembly-replication intermediates were imaged using electron microscopy

Electron microscopic analysis of rotavirus assembly-replication intermediates

Energy Technology Data Exchange (ETDEWEB)

Boudreaux, Crystal E.; Kelly, Deborah F. [Virginia Tech Carilion School of Medicine and Research Institute, Roanoke, VA (United States); McDonald, Sarah M., E-mail: mcdonaldsa@vtc.vt.edu [Virginia Tech Carilion School of Medicine and Research Institute, Roanoke, VA (United States); Department of Biomedical Sciences and Pathobiology, Virginia—Maryland Regional College of Veterinary Medicine, Blacksburg, VA (United States)

2015-03-15

Rotaviruses (RVs) replicate their segmented, double-stranded RNA genomes in tandem with early virion assembly. In this study, we sought to gain insight into the ultrastructure of RV assembly-replication intermediates (RIs) using transmission electron microscopy (EM). Specifically, we examined a replicase-competent, subcellular fraction that contains all known RV RIs. Three never-before-seen complexes were visualized in this fraction. Using in vitro reconstitution, we showed that ~15-nm doughnut-shaped proteins in strings were nonstructural protein 2 (NSP2) bound to viral RNA transcripts. Moreover, using immunoaffinity-capture EM, we revealed that ~20-nm pebble-shaped complexes contain the viral RNA polymerase (VP1) and RNA capping enzyme (VP3). Finally, using a gel purification method, we demonstrated that ~30–70-nm electron-dense, particle-shaped complexes represent replicase-competent core RIs, containing VP1, VP3, and NSP2 as well as capsid proteins VP2 and VP6. The results of this study raise new questions about the interactions among viral proteins and RNA during the concerted assembly–replicase process. - Highlights: • Rotaviruses replicate their genomes in tandem with early virion assembly. • Little is known about rotavirus assembly-replication intermediates. • Assembly-replication intermediates were imaged using electron microscopy.

Assembling large, complex environmental metagenomes

Energy Technology Data Exchange (ETDEWEB)

Howe, A. C. [Michigan State Univ., East Lansing, MI (United States). Microbiology and Molecular Genetics, Plant Soil and Microbial Sciences; Jansson, J. [USDOE Joint Genome Institute (JGI), Walnut Creek, CA (United States); Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). Earth Sciences Division; Malfatti, S. A. [USDOE Joint Genome Institute (JGI), Walnut Creek, CA (United States); Tringe, S. G. [USDOE Joint Genome Institute (JGI), Walnut Creek, CA (United States); Tiedje, J. M. [Michigan State Univ., East Lansing, MI (United States). Microbiology and Molecular Genetics, Plant Soil and Microbial Sciences; Brown, C. T. [Michigan State Univ., East Lansing, MI (United States). Microbiology and Molecular Genetics, Computer Science and Engineering

2012-12-28

The large volumes of sequencing data required to sample complex environments deeply pose new challenges to sequence analysis approaches. De novo metagenomic assembly effectively reduces the total amount of data to be analyzed but requires significant computational resources. We apply two pre-assembly filtering approaches, digital normalization and partitioning, to make large metagenome assemblies more computationaly tractable. Using a human gut mock community dataset, we demonstrate that these methods result in assemblies nearly identical to assemblies from unprocessed data. We then assemble two large soil metagenomes from matched Iowa corn and native prairie soils. The predicted functional content and phylogenetic origin of the assembled contigs indicate significant taxonomic differences despite similar function. The assembly strategies presented are generic and can be extended to any metagenome; full source code is freely available under a BSD license.

Criticality calculations of various spent fuel casks - possibilities for burn up credit implementation

International Nuclear Information System (INIS)

Apostolov, T; Manolova, M.; Prodanova, R.

2001-01-01

A methodology for criticality safety analysis of spent fuel casks with possibilities for burnup credit implementation is presented. This methodology includes the world well-known and applied program systems: NESSEL-NUKO for depletion and SCALE-4.4 for criticality calculations. The abilities of this methodology to analyze storage and transportation casks with different type of spent fuel are demonstrated on the base of various tests. The depletion calculations have been carried out for the power reactors (WWER-440 and WWER-1000) and the research reactor IRT-2000 (C-36) fuel assemblies. The criticality calculation models have been developed on the basis of real fuel casks, designed by the leading international companies (for WWER-440 and WWER-1000 spent fuel assemblies), as well as for real a WWER-440 storage cask, applied at the 'Kozloduy' NPP. The results obtained show that the criticality safety criterion K eff less than 0.95 is satisfied for both: fresh and spent fuel. Besides the implementation of burnup credit allows to account for the reduced reactivity of spent fuel and to evaluate the conservatism of the fresh fuel assumption. (author)

Development of Mathematical Model and Analysis Code for Estimating Drop Behavior of the Control Rod Assembly in the Sodium Cooled Fast Reactor

International Nuclear Information System (INIS)

Oh, Se-Hong; Kang, SeungHoon; Choi, Choengryul; Yoon, Kyung Ho; Cheon, Jin Sik

2016-01-01

On receiving the scram signal, the control rod assemblies are released to fall into the reactor core by its weight. Thus drop time and falling velocity of the control rod assembly must be estimated for the safety evaluation. There are three typical ways to estimate the drop behavior of the control rod assembly in scram action: Experimental, numerical and theoretical methods. But experimental and numerical(CFD) method require a lot of cost and time. Thus, these methods are difficult to apply to the initial design process. In this study, mathematical model and theoretical analysis code have been developed in order to estimate drop behavior of the control rod assembly to provide the underlying data for the design optimization. Mathematical model and theoretical analysis code have been developed in order to estimate drop behavior of the control rod assembly to provide the underlying data for the design optimization. A simplified control rod assembly model is considered to minimize the uncertainty in the development process. And the hydraulic circuit analysis technique is adopted to evaluate the internal/external flow distribution of the control rod assembly. Finally, the theoretical analysis code(named as HEXCON) has been developed based on the mathematical model. To verify the reliability of the developed code, CFD analysis has been conducted. And a calculation using the developed analysis code was carried out under the same condition, and both results were compared

Critical feature analysis of a radiotherapy machine

International Nuclear Information System (INIS)

Rae, Andrew; Jackson, Daniel; Ramanan, Prasad; Flanz, Jay; Leyman, Didier

2005-01-01

The software implementation of the emergency shutdown feature in a major radiotherapy system was analyzed, using a directed form of code review based on module dependences. Dependences between modules are labelled by particular assumptions; this allows one to trace through the code, and identify those fragments responsible for critical features. An 'assumption tree' is constructed in parallel, showing the assumptions which each module makes about others. The root of the assumption tree is the critical feature of interest, and its leaves represent assumptions which, if not valid, might cause the critical feature to fail. The analysis revealed some unexpected assumptions that motivated improvements to the code

21 CFR 120.8 - Hazard Analysis and Critical Control Point (HACCP) plan.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 2 2010-04-01 2010-04-01 false Hazard Analysis and Critical Control Point (HACCP... SERVICES (CONTINUED) FOOD FOR HUMAN CONSUMPTION HAZARD ANALYSIS AND CRITICAL CONTROL POINT (HACCP) SYSTEMS General Provisions § 120.8 Hazard Analysis and Critical Control Point (HACCP) plan. (a) HACCP plan. Each...

Criticality safety considerations in the geologic disposal of spent nuclear fuel assemblies

International Nuclear Information System (INIS)

Gore, B.F.; McNair, G.W.; Heaberlin, S.W.

1980-05-01

Features of geologic disposal which hamper the demonstration that criticality cannot occur therein include possible changes of shape and form, intrusion of water as a neutron moderator, and selective leaching of spent fuel constituents. If the criticality safety of spent fuel disposal depends on burnup, independent measurements verifying the burnup should be performed prior to disposal. The status of nondestructive analysis method which might provide such verification is discussed. Calculations were performed to assess the potential for increasing the allowed size of a spent fuel disposal canister if potential water intrusion were limited by close-packing the enclosed rods. Several factors were identified which severely limited the potential of this application. The theoretical limit of hexagonal close-packing cannot be achieved due to fuel rod bowing. It is concluded that disposal canisters should be sized on the basis of assumed optimum moderation. Several topics for additional research were identified during this limited study

Challenges in the vulnerability and risk analysis of critical infrastructures

International Nuclear Information System (INIS)

Zio, Enrico

2016-01-01

The objective of this paper is to provide a systematic view on the problem of vulnerability and risk analysis of critical infrastructures. Reflections are made on the inherent complexities of these systems, related challenges are identified and possible ways forward for their analysis and management are indicated. Specifically: the framework of vulnerability and risk analysis is examined in relation to its application for the protection and resilience of critical infrastructures; it is argued that the complexity of these systems is a challenging characteristic, which calls for the integration of different modeling perspectives and new approaches of analysis; examples of are given in relation to the Internet and, particularly, the electric power grid, as representative of critical infrastructures and the associated complexity; the integration of different types of analyses and methods of system modeling is put forward for capturing the inherent structural and dynamic complexities of critical infrastructures and eventually evaluating their vulnerability and risk characteristics, so that decisions on protections and resilience actions can be taken with the required confidence. - Highlights: • The problem of the protection and resilience of CIs is the focus of the work. • The vulnerability and risk analysis framework for this is critically examined. • The complexity of CIs is presented as a challenge for system modeling and analysis. • The integration of different modeling perspectives of analysis is put forward as a solution. • The extension of the analysis framework to new methods for dealing with surprises and black swans is advocated.

Burn-up Credit Criticality Safety Benchmark Phase III-C. Nuclide Composition and Neutron Multiplication Factor of a Boiling Water Reactor Spent Fuel Assembly for Burn-up Credit and Criticality Control of Damaged Nuclear Fuel

International Nuclear Information System (INIS)

Suyama, K.; Uchida, Y.; Kashima, T.; Ito, T.; Miyaji, T.

2016-01-01

Criticality control of damaged nuclear fuel is one of the key issues in the decommissioning operation of the Fukushima Daiichi Nuclear Power Station accident. The average isotopic composition of spent nuclear fuel as a function of burn-up is required in order to evaluate criticality parameters of the mixture of damaged nuclear fuel with other materials. The NEA Expert Group on Burn-up Credit Criticality (EGBUC) has organised several international benchmarks to assess the accuracy of burn-up calculation methodologies. For BWR fuel, the Phase III-B benchmark, published in 2002, was a remarkable landmark that provided general information on the burn-up properties of BWR spent fuel based on the 8x8 type fuel assembly. Since the publication of the Phase III-B benchmark, all major nuclear data libraries have been revised; in Japan from JENDL-3.2 to JENDL-4, in Europe from JEF-2.2 to JEFF-3.1 and in the US from ENDF/B-VI to ENDF/B-VII.1. Burn-up calculation methodologies have been improved by adopting continuous-energy Monte Carlo codes and modern neutronics calculation methods. Considering the importance of the criticality control of damaged fuel in the Fukushima Daiichi Nuclear Power Station accident, a new international burn-up calculation benchmark for the 9 x 9 STEP-3 BWR fuel assemblies was organised to carry out the inter-comparison of the averaged isotopic composition in the interest of the burnup credit criticality safety community. Benchmark specifications were proposed and approved at the EGBUC meeting in September 2012 and distributed in October 2012. The deadline for submitting results was set at the end of February 2013. The basic model for the benchmark problem is an infinite two-dimensional array of BWR fuel assemblies consisting of a 9 x 9 fuel rod array with a water channel in the centre. The initial uranium enrichment of fuel rods without gadolinium is 4.9, 4.4, 3.9, 3.4 and 2.1 wt% and 3.4 wt% for the rods using gadolinium. The burn-up conditions are

Sequence Identification, Recombinant Production, and Analysis of the Self-Assembly of Egg Stalk Silk Proteins from Lacewing Chrysoperla carnea.

Science.gov (United States)

Neuenfeldt, Martin; Scheibel, Thomas

2017-06-13

Egg stalk silks of the common green lacewing Chrysoperla carnea likely comprise at least three different silk proteins. Based on the natural spinning process, it was hypothesized that these proteins self-assemble without shear stress, as adult lacewings do not use a spinneret. To examine this, the first sequence identification and determination of the gene expression profile of several silk proteins and various transcript variants thereof was conducted, and then the three major proteins were recombinantly produced in Escherichia coli encoded by their native complementary DNA (cDNA) sequences. Circular dichroism measurements indicated that the silk proteins in aqueous solutions had a mainly intrinsically disordered structure. The largest silk protein, which we named ChryC1, exhibited a lower critical solution temperature (LCST) behavior and self-assembled into fibers or film morphologies, depending on the conditions used. The second silk protein, ChryC2, self-assembled into nanofibrils and subsequently formed hydrogels. Circular dichroism and Fourier transform infrared spectroscopy confirmed conformational changes of both proteins into beta sheet rich structures upon assembly. ChryC3 did not self-assemble into any morphology under the tested conditions. Thereby, through this work, it could be shown that recombinant lacewing silk proteins can be produced and further used for studying the fiber formation of lacewing egg stalks.

Directed Self-Assembly of Nanodispersions

Energy Technology Data Exchange (ETDEWEB)

Furst, Eric M [University of Delaware

2013-11-15

Directed self-assembly promises to be the technologically and economically optimal approach to industrial-scale nanotechnology, and will enable the realization of inexpensive, reproducible and active nanostructured materials with tailored photonic, transport and mechanical properties. These new nanomaterials will play a critical role in meeting the 21st century grand challenges of the US, including energy diversity and sustainability, national security and economic competitiveness. The goal of this work was to develop and fundamentally validate methods of directed selfassembly of nanomaterials and nanodispersion processing. The specific aims were: 1. Nanocolloid self-assembly and interactions in AC electric fields. In an effort to reduce the particle sizes used in AC electric field self-assembly to lengthscales, we propose detailed characterizations of field-driven structures and studies of the fundamental underlying particle interactions. We will utilize microscopy and light scattering to assess order-disorder transitions and self-assembled structures under a variety of field and physicochemical conditions. Optical trapping will be used to measure particle interactions. These experiments will be synergetic with calculations of the particle polarizability, enabling us to both validate interactions and predict the order-disorder transition for nanocolloids. 2. Assembly of anisotropic nanocolloids. Particle shape has profound effects on structure and flow behavior of dispersions, and greatly complicates their processing and self-assembly. The methods developed to study the self-assembled structures and underlying particle interactions for dispersions of isotropic nanocolloids will be extended to systems composed of anisotropic particles. This report reviews several key advances that have been made during this project, including, (1) advances in the measurement of particle polarization mechanisms underlying field-directed self-assembly, and (2) progress in the

Micellar Self-Assembly of Recombinant Resilin-/Elastin-Like Block Copolypeptides.

Science.gov (United States)

Weitzhandler, Isaac; Dzuricky, Michael; Hoffmann, Ingo; Garcia Quiroz, Felipe; Gradzielski, Michael; Chilkoti, Ashutosh

2017-08-14

Reported here is the synthesis of perfectly sequence defined, monodisperse diblock copolypeptides of hydrophilic elastin-like and hydrophobic resilin-like polypeptide blocks and characterization of their self-assembly as a function of structural parameters by light scattering, cryo-TEM, and small-angle neutron scattering. A subset of these diblock copolypeptides exhibit lower critical solution temperature and upper critical solution temperature phase behavior and self-assemble into spherical or cylindrical micelles. Their morphologies are dictated by their chain length, degree of hydrophilicity, and hydrophilic weight fraction of the ELP block. We find that (1) independent of the length of the corona-forming ELP block there is a minimum threshold in the length of the RLP block below which self-assembly does not occur, but that once that threshold is crossed, (2) the RLP block length is a unique molecular parameter to independently tune self-assembly and (3) increasing the hydrophobicity of the corona-forming ELP drives a transition from spherical to cylindrical morphology. Unlike the self-assembly of purely ELP-based block copolymers, the self-assembly of RLP-ELPs can be understood by simple principles of polymer physics relating hydrophilic weight fraction and polymer-polymer and polymer-solvent interactions to micellar morphology, which is important as it provides a route for the de novo design of desired nanoscale morphologies from first principles.

Development of fuel assembly seismic analysis against vertical and horizontal earthquake

International Nuclear Information System (INIS)

Sato, T.; Akitake, J.; Kobayashi, H.; Azumi, S.; Koike, H.; Takeda, N.; Suzuki, S.

2001-01-01

Vertical vibration with large acceleration was observed in KOBE earthquake in 1995. Concerning PWR fuel assembly, though the vertical response has been calculated by a static analysis, it had better be calculated by a dynamic analysis in detail. Furthermore, mutual effects between horizontal and vertical motions attract our attention. For these reasons, a dynamic analysis method in the vertical direction was developed and linked with the previously developed method in the horizontal direction. This is the method that takes effect of vertical vibration into the horizontal vibration analysis as the change of horizontal stiffness, which is brought by axial compressive force. In this paper, fundamental test results for developing the method are introduced and summary of the advanced method's procedure and analysis results are also described. (authors)

Spatio-temporal Dynamics and Mechanisms of Stress Granule Assembly.

Directory of Open Access Journals (Sweden)

Daisuke Ohshima

2015-06-01

Full Text Available Stress granules (SGs are non-membranous cytoplasmic aggregates of mRNAs and related proteins, assembled in response to environmental stresses such as heat shock, hypoxia, endoplasmic reticulum (ER stress, chemicals (e.g. arsenite, and viral infections. SGs are hypothesized as a loci of mRNA triage and/or maintenance of proper translation capacity ratio to the pool of mRNAs. In brain ischemia, hippocampal CA3 neurons, which are resilient to ischemia, assemble SGs. In contrast, CA1 neurons, which are vulnerable to ischemia, do not assemble SGs. These results suggest a critical role SG plays in regards to cell fate decisions. Thus SG assembly along with its dynamics should determine the cell fate. However, the process that exactly determines the SG assembly dynamics is largely unknown. In this paper, analyses of experimental data and computer simulations were used to approach this problem. SGs were assembled as a result of applying arsenite to HeLa cells. The number of SGs increased after a short latent period, reached a maximum, then decreased during the application of arsenite. At the same time, the size of SGs grew larger and became localized at the perinuclear region. A minimal mathematical model was constructed, and stochastic simulations were run to test the modeling. Since SGs are discrete entities as there are only several tens of them in a cell, commonly used deterministic simulations could not be employed. The stochastic simulations replicated observed dynamics of SG assembly. In addition, these stochastic simulations predicted a gamma distribution relative to the size of SGs. This same distribution was also found in our experimental data suggesting the existence of multiple fusion steps in the SG assembly. Furthermore, we found that the initial steps in the SG assembly process and microtubules were critical to the dynamics. Thus our experiments and stochastic simulations presented a possible mechanism regulating SG assembly.

Criticality Safety Evaluation for the TACS at DAF

Energy Technology Data Exchange (ETDEWEB)

Percher, C. M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Heinrichs, D. P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2011-06-10

Hands-on experimental training in the physical behavior of multiplying systems is one of ten key areas of training required for practitioners to become qualified in the discipline of criticality safety as identified in DOE-STD-1135-99, Guidance for Nuclear Criticality Safety Engineer Training and Qualification. This document is a criticality safety evaluation of the training activities and operations associated with HS-3201-P, Nuclear Criticality 4-Day Training Course (Practical). This course was designed to also address the training needs of nuclear criticality safety professionals under the auspices of the NNSA Nuclear Criticality Safety Program1. The hands-on, or laboratory, portion of the course will utilize the Training Assembly for Criticality Safety (TACS) and will be conducted in the Device Assembly Facility (DAF) at the Nevada Nuclear Security Site (NNSS). The training activities will be conducted by Lawrence Livermore National Laboratory following the requirements of an Integrated Work Sheet (IWS) and associated Safety Plan. Students will be allowed to handle the fissile material under the supervision of an LLNL Certified Fissile Material Handler.

Searching for scientific literacy and critical pedagogy in socioscientific curricula: A critical discourse analysis

Science.gov (United States)

Cummings, Kristina M.

The omnipresence of science and technology in our society require the development of a critical and scientifically literate citizenry. However, the inclusion of socioscientific issues, which are open-ended controversial issues informed by both science and societal factors such as politics, economics, and ethics, do not guarantee the development of these skills. The purpose of this critical discourse analysis is to identify and analyze the discursive strategies used in intermediate science texts and curricula that address socioscientific topics and the extent to which the discourses are designed to promote or suppress the development of scientific literacy and a critical pedagogy. Three curricula that address the issue of energy and climate change were analyzed using Gee's (2011) building tasks and inquiry tools. The curricula were written by an education organization entitled PreSEES, a corporate-sponsored group called NEED, and a non-profit organization named Oxfam. The analysis found that the PreSEES and Oxfam curricula elevated the significance of climate change and the NEED curriculum deemphasized the issue. The PreSEES and Oxfam curricula promoted the development of scientific literacy while the NEED curricula suppressed its development. The PreSEES and Oxfam curricula both promoted the development of the critical pedagogy; however, only the Oxfam curricula provided authentic opportunities to enact sociopolitical change. The NEED curricula suppressed the development of critical pedagogy. From these findings, the following conclusions were drawn. When socioscientific issues are presented with the development of scientific literacy and critical pedagogy, the curricula allow students to develop fact-based opinions about the issue. However, curricula that address socioscientific issues without the inclusion of these skills minimize the significance of the issue and normalize the hegemonic worldview promoted by the curricula's authors. Based on these findings

Development of an Advanced Recycle Filter Tank Assembly for the ISS Urine Processor Assembly

Science.gov (United States)

Link, Dwight E., Jr.; Carter, Donald Layne; Higbie, Scott

2010-01-01

Recovering water from urine is a process that is critical to supporting larger crews for extended missions aboard the International Space Station. Urine is collected, preserved, and stored for processing into water and a concentrated brine solution that is highly toxic and must be contained to avoid exposure to the crew. The brine solution is collected in an accumulator tank, called a Recycle Filter Tank Assembly (RFTA) that must be replaced monthly and disposed in order to continue urine processing operations. In order to reduce resupply requirements, a new accumulator tank is being developed that can be emptied on orbit into existing ISS waste tanks. The new tank, called the Advanced Recycle Filter Tank Assembly (ARFTA) is a metal bellows tank that is designed to collect concentrated brine solution and empty by applying pressure to the bellows. This paper discusses the requirements and design of the ARFTA as well as integration into the urine processor assembly.

Criticality safety analysis for plutonium dissolver using silver mediated electrolytic oxidation method

International Nuclear Information System (INIS)

Umeda, Miki; Sugikawa, Susumu; Nakamura, Kazuhito; Egashira, Tetsurou

1998-08-01

Design and construction of a plutonium dissolver using silver mediated electrolytic oxidation method are promoted in NUCEF. Criticality safety analysis for the plutonium dissolver is described in this report. The electrolytic plutonium dissolver consists of connection pipes and three pots for MOX powder supply, circulation and electrolysis. The criticality control for the dissolver is made by geometrically safe shape with mass limitation. Monte Carlo code KENO-IV using MGCL-137 library based on ENDF/B-IV was used for the criticality safety analysis for the plutonium dissolver. Considering the required size for construction and criticality safety, diameter of pot and distance between two pots were determined. On this condition, the criticality safety analysis for the plutonium dissolver with connection pipes was carried out. As the result of the criticality safety analysis, an effective neutron multiplication factor keff of 0.91 was obtained and the criticality safety of the plutonium dissolver was confirmed on the basis of criteria of ≤0.95. (author)

Toward superconducting critical current by design

OpenAIRE

Sadovskyy, I. A.; Jia, Y.; Leroux, M.; Kwon, J.; Hu, H.; Fang, L.; Chaparro, C.; Zhu, S.; Welp, U.; Zuo, J. -M.; Zhang, Y.; Nakasaki, R.; Selvamanickam, V.; Crabtree, G. W.; Koshelev, A. E.

2015-01-01

We present the new paradigm of critical current by design. Analogous to materials by design, it aims at predicting the optimal defect landscape in a superconductor for targeted applications by elucidating the vortex dynamics responsible for the bulk critical current. To highlight this approach, we demonstrate the synergistic combination of critical current measurements on commercial high-temperature superconductors containing self-assembled and irradiation tailored correlated defects by using...

Fuel assembly stress and deflection analysis for loss-of-coolant accident and seismic excitation

International Nuclear Information System (INIS)

DeMars, R.V.; Steinke, R.R.

1975-01-01

Babcock and Wilcox has evaluated the capability of the fuel assemblies to withstand the effects of a loss-of-coolant accident (LOCA) blowdown, the operational basis earthquake (OBE) and design basis earthquake (DBE), and the simultaneous occurrence of the DBE and LOCA. This method of analysis is applicable to all of B and W's nuclear steam system contracts that specify the skirt-supported pressure vessel. Loads during the saturated and subcooled phases of blowdown following a loss-of-coolant accident were calculated. The maximum loads on the fuel assemblies were found to be below allowable limits, and the maximum deflections of the fuel assemblies were found to be less than those that could prevent the insertion of control rods or the flow of coolant through the core. (U.S.)

Verification and sensitivity analysis on the elastic stiffness of the leaf type holddown spring assembly

International Nuclear Information System (INIS)

Song, Kee Nam

1998-01-01

The elastic formula of leaf type hold down spring(HDS) assembly is verified by comparing the values of elastic stiffness with the characteristic test results of the HDS's specimens. The comparisons show that the derived elastic stiffness formula is useful in reliably estimating the elastic stiffness of leaf type HDS assembly. The elastic stiffness sensitivity of leaf type HDS assembly is analyzed using the formula and its gradient vectors obtained from the mid-point formula. As a result of sensitivity analysis, the elastic stiffness sensitivity with respect to each design variable is quantified and design variables of large sensitivity are identified. Among the design variables, leaf thickness is identified as the most sensitive design variable to the elastic of leaf type HDS assembly. In addition, the elastic stiffness sensitivity, with respect to design variable, is in power-law type correlation to the base thickness of the leaf. (author)

Cognitive systems engineering analysis of the JCO criticality accident

International Nuclear Information System (INIS)

Tanabe, Fumiya; Yamaguchi, Yukichi

2000-01-01

The JCO Criticality Accident is analyzed with a framework based on cognitive systems engineering. With the framework, analysis is conducted integrally both from the system viewpoint and actors viewpoint. The occupational chemical risk was important as safety constraint for the actors as well as the nuclear risk, which is due to criticality accident, to the public and to actors. The inappropriate actor's mental model of the work system played a critical role and several factors (e.g. poor training and education, lack of information on criticality safety control in the procedures and instructions, and lack of warning signs at workplace) contributed to form and shape the mental model. Based on the analysis, several countermeasures, such as warning signs, information system for supporting actors and improved training and education, are derived to prevent such an accident. (author)

Analysis of strain distribution and critical current of superconductors based on a strain-critical current measurement system

International Nuclear Information System (INIS)

Liu Fang; Wu Yu; Long Feng

2010-01-01

Based on Pacman device which is widely used to investigate the axial strain dependence of the critical current in superconductors, the finite element analysis method is employed to carry out the force analysis of the spring and the superconducting strand, thereby the axial and lateral strain distributions of the superconducting strand are obtained. According to the two extreme assumptions(low inter-filament resistance and high inter-filament resistance), the effects of the strain homogeneity at the cross section of the superconductor on the critical current is analyzed combined with the Nb 3 Sn deviatoric strain-critical current scaling law. (authors)

Criticality Calculations of Fresh LEU and MOX Assemblies for Transport and Storage at the Balakovo Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Goluoglu, S.

2001-01-11

Transportation of low-enriched uranium (LEU) and mixed-oxide (MOX) assemblies to and within the VVER-1000-type Balakovo Nuclear Power Plant is investigated. Effective multiplication factors for fresh fuel assemblies on the railroad platform, fresh fuel assemblies in the fuel transportation vehicle, and fresh fuel assemblies in the spent fuel storage pool are calculated. If there is no absorber between the units, the configurations with all MOX assemblies result in higher effective multiplication factors than the configurations with all LEU assemblies when the system is dry. When the system is flooded, the configurations with all LEU assemblies result in higher effective multiplication factors. For normal operating conditions, effective multiplication factors for all configurations are below the presumed upper subcritical limit of 0.95. For an accident condition of a fully loaded fuel transportation vehicle that is flooded with low-density water (possibly from a fire suppression system), the presumed upper subcritical limit is exceeded by configurations containing LEU assemblies.

Genome Sequence Databases (Overview): Sequencing and Assembly

Energy Technology Data Exchange (ETDEWEB)

Lapidus, Alla L.

2009-01-01

From the date its role in heredity was discovered, DNA has been generating interest among scientists from different fields of knowledge: physicists have studied the three dimensional structure of the DNA molecule, biologists tried to decode the secrets of life hidden within these long molecules, and technologists invent and improve methods of DNA analysis. The analysis of the nucleotide sequence of DNA occupies a special place among the methods developed. Thanks to the variety of sequencing technologies available, the process of decoding the sequence of genomic DNA (or whole genome sequencing) has become robust and inexpensive. Meanwhile the assembly of whole genome sequences remains a challenging task. In addition to the need to assemble millions of DNA fragments of different length (from 35 bp (Solexa) to 800 bp (Sanger)), great interest in analysis of microbial communities (metagenomes) of different complexities raises new problems and pushes some new requirements for sequence assembly tools to the forefront. The genome assembly process can be divided into two steps: draft assembly and assembly improvement (finishing). Despite the fact that automatically performed assembly (or draft assembly) is capable of covering up to 98% of the genome, in most cases, it still contains incorrectly assembled reads. The error rate of the consensus sequence produced at this stage is about 1/2000 bp. A finished genome represents the genome assembly of much higher accuracy (with no gaps or incorrectly assembled areas) and quality ({approx}1 error/10,000 bp), validated through a number of computer and laboratory experiments.

Toward high throughput optical metamaterial assemblies.

Science.gov (United States)

Fontana, Jake; Ratna, Banahalli R

2015-11-01

Optical metamaterials have unique engineered optical properties. These properties arise from the careful organization of plasmonic elements. Transitioning these properties from laboratory experiments to functional materials may lead to disruptive technologies for controlling light. A significant issue impeding the realization of optical metamaterial devices is the need for robust and efficient assembly strategies to govern the order of the nanometer-sized elements while enabling macroscopic throughput. This mini-review critically highlights recent approaches and challenges in creating these artificial materials. As the ability to assemble optical metamaterials improves, new unforeseen opportunities may arise for revolutionary optical devices.

Criticality safety of spent fuel casks considering water inleakage

International Nuclear Information System (INIS)

Osgood, N.L.; Withee, C.J.; Easton, E.P.

2004-01-01

A fundamental safety design parameter for all fissile material packages is that a single package must be critically safe even if water leaks into the containment system. In addition, criticality safety must be assured for arrays of packages under normal conditions of transport (undamaged packages) and under hypothetical accident conditions (damaged packages). The U.S. Nuclear Regulatory Commission staff has revised the review protocol for demonstrating criticality safety for spent fuel casks. Previous review guidance specified that water inleakage be considered under accident conditions. This practice was based on the fact that the leak tightness of spent fuel casks is typically demonstrated by use of structural analysis and not by physical testing. In addition, since a single package was shown to be safe with water inleakage, it was concluded that this analysis was also applicable to an array of damaged packages, since the heavy shield walls in spent fuel casks neutronically isolate each cask in the array. Inherent in this conclusion is that the fuel assembly geometry does not change significantly, even under drop test conditions. Requests for shipping fuel with burnup exceeding 40 GWd/MTU, including very high burnups exceeding 60 GWD/MTU, caused a reassessment of this assumption. Fuel cladding structural strength and ductility were not clearly predictable for these higher burnups. Therefore the single package analysis for an undamaged package may not be applicable for the damaged package. NRC staff developed a new practice for review of spent fuel casks under accident conditions. The practice presents two methods for approval that would allow an assessment of potential reconfiguration of the fuel assembly under accident conditions, or, alternatively, a demonstration of the water-exclusion boundary through physical testing

Criticality analysis in uranium enrichment plant

International Nuclear Information System (INIS)

Okamoto, Tsuyoshi; Kiyose, Ryohei

1977-01-01

In a large scale uranium enrichment plant, uranium inventory in cascade rooms is not very large in quantity, but the facilities dealing with the largest quantity of uranium in that process are the UF 6 gas supply system and the blending system for controlling the product concentration. When UF 6 spills out of these systems, the enriched uranium is accumulated, and the danger of criticality accident is feared. If a NaF trap is placed at the forestage of waste gas treatment system, plenty of UF 6 and HF are adsorbed together in the NaF trap. Thus, here is the necessity of checking the safety against criticality. Various assumptions were made to perform the computation surveying the criticality of the system composed of UF 6 and HF adsorbed on NaF traps with WIMS code (transport analysis). The minimum critical radius resulted in about 53 cm in case of 3.5% enriched fuel for light water reactors. The optimum volume ratio of fissile material in the double salt UF 6 .2NaF and NaF.HF is about 40 vol. %. While, criticality survey computation was also made for the annular NaF trap having the central cooling tube, and it was found that the effect of cooling tube radius did not decrease the multiplication factor up to the cooling tube radius of about 5 cm. (Wakatsuki, Y.)

α-Cut method based importance measure for criticality analysis in fuzzy probability – Based fault tree analysis

International Nuclear Information System (INIS)

Purba, Julwan Hendry; Sony Tjahyani, D.T.; Widodo, Surip; Tjahjono, Hendro

2017-01-01

Highlights: •FPFTA deals with epistemic uncertainty using fuzzy probability. •Criticality analysis is important for reliability improvement. •An α-cut method based importance measure is proposed for criticality analysis in FPFTA. •The α-cut method based importance measure utilises α-cut multiplication, α-cut subtraction, and area defuzzification technique. •Benchmarking confirm that the proposed method is feasible for criticality analysis in FPFTA. -- Abstract: Fuzzy probability – based fault tree analysis (FPFTA) has been recently developed and proposed to deal with the limitations of conventional fault tree analysis. In FPFTA, reliabilities of basic events, intermediate events and top event are characterized by fuzzy probabilities. Furthermore, the quantification of the FPFTA is based on fuzzy multiplication rule and fuzzy complementation rule to propagate uncertainties from basic event to the top event. Since the objective of the fault tree analysis is to improve the reliability of the system being evaluated, it is necessary to find the weakest path in the system. For this purpose, criticality analysis can be implemented. Various importance measures, which are based on conventional probabilities, have been developed and proposed for criticality analysis in fault tree analysis. However, not one of those importance measures can be applied for criticality analysis in FPFTA, which is based on fuzzy probability. To be fully applied in nuclear power plant probabilistic safety assessment, FPFTA needs to have its corresponding importance measure. The objective of this study is to develop an α-cut method based importance measure to evaluate and rank the importance of basic events for criticality analysis in FPFTA. To demonstrate the applicability of the proposed measure, a case study is performed and its results are then benchmarked to the results generated by the four well known importance measures in conventional fault tree analysis. The results

The initial criticality and nuclear commissioning test program at HANARO

Energy Technology Data Exchange (ETDEWEB)

Lee, Choong-Sung; Seo, Chul-Gyo; Jun, Byung-Jin [Korea Atomic Energy Research Institute, Dukjin-Dong 150, Yusung-Ku, Taejon, 305-353 (Korea, Republic of)

1995-07-01

The construction of the Korea Multipurpose Research Reactor - HANARO of 3MW, developed by Korea Atomic Energy Research Institute, was completed at the beginning of this year. The first fuel loading began on February 2 1995, and initial criticality was achieved on February 8, when the core had four 18-element assemblies and thirteen 36-element assemblies. The critical control rod position was 600.8 mm which represents excess reactivity of 0.71 $. Currently the nuclear commissioning test is on going under the zero power range. This paper describes the initial criticality approach of the HANARO, and its nuclear commissioning test program. (author)

Status and future program of reactor physics experiments in JAERI Critical facilities, FCA and TCA

International Nuclear Information System (INIS)

Okajima, Shigeaki; Osugi, Toshitaka; Nakajima, Ken; Suzaki, Takenori; Miyoshi, Yoshinori

1999-01-01

The critical facilities in JAERI, FCA (Fast Critical Assembly) and TCA (Tank-type Critical Assembly), have been used to provide integral data for evaluation of nuclear data as well as for development of various types of reactor since they went critical in 1960's. In this paper a review is presented on the experimental programs in both facilities. And the experimental programs in next 5 years are also shown. (author)

Validating analysis methodologies used in burnup credit criticality calculations

International Nuclear Information System (INIS)

Brady, M.C.; Napolitano, D.G.

1992-01-01