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Forecast of criticality experiments and experimental programs needed to support nuclear operations in the United States of America: 1994--1999
Abstract
This Forecast is generated by the Chair of the Experiment Needs Identification Workgroup (ENIWG), with input from Department of Energy and the nuclear community. One of the current concerns addressed by ENIWG was the Defense Nuclear Facilities Safety Board`s Recommendation 93-2. This Recommendation delineated the need for a critical experimental capability, which includes (1) a program of general-purpose experiments, (2) improving the information base, and (3) ongoing departmental programs. The nuclear community also recognizes the importance of criticality theory, which, as a stepping stone to computational analysis and safety code development, needs to be benchmarked against well-characterized critical experiments. A summary project of the Department`s needs with respect to criticality information includes (1) hands-on training, (2) criticality and nuclear data, (3) detector systems, (4) uranium- and plutonium-based reactors, and (5) accident analysis. The Workgroup has evaluated, prioritized, and categorized each proposed experiment and program. Transportation/Applications is a new category intended to cover the areas of storage, training, emergency response, and standards. This category has the highest number of priority-1 experiments (nine). Facilities capable of performing experiments include the Los Alamos Critical Experiment Facility (LACEF) along with Area V at Sandia National Laboratory. The LACEF continues to house the most significant collection of critical assemblies in the Western Hemisphere. The staff of this facility and Area V are trained and certified, and documentation is current. ENIWG will continue to work with the nuclear community to identify and prioritize experiments because there is an overwhelming need for critical experiments to be performed for basic research and code validation.
... In May 1998, the chair of the Experiment Needs Identification Workgroup ranked the need for experiments with graphite, Be, BeO, and D 2 O in the top ten in the revised recommendations for priority of critical experiments. 15 These experiments were identified as "Special Moderator Experiments," and they were next in the queue for experiments to be formed at the LACEF prior to the shutdown and subsequent move of critical assemblies to the Device Assembly Facility in Las Vegas, Nevada. Graphite has a large number of resonances for the scattering cross sections in the intermediate energy region, and little or no experimental data exist in this energy region. ...
The National Aeronautics and Space Administration is considering nuclear power sources for space exploration. A series of critical mass experiments was designed to address the development, performance, and design of a space nuclear reactor being considered to support the Prometheus project. These experiments consisted of interlacing the refractory metals rhenium (Re), molybdenum (Mo), tantalum2.5 wt% tungsten (Ta-2.5W), and niobium-1 wt% zirconium (Nb-1Zr) with moderating materials (graphite or polyethylene) and were fueled by highly enriched uranium plates. These experiments are designed to assess the adequacy of and uncertainty in refractory metal neutron cross-section evaluations for use in Prometheus nuclear reactor design work. The critical experiments were designed in the energy spectrum closely resembling or bracketing that in the proposed space reactor. For each material (Re, Mo, Ta-2.5W and Nb-1Zr), four critical configurations were designed and performed to measure the sensitivity of keff to the material under four different and progressively softer neutron spectra (core center spectrum, harder than core average spectrum, softer than core average spectrum, and accident flooded spectrum). The thicknesses of the graphite or polyethylene moderator and reflector plates were adjusted to achieve the desired neutron spectrum. One critical and 18 subcritical experiments provided for measurements of material neutronic behavior in a simple cylindrical geometry configuration that was modeled in MCNP with ENDF/B-VI.6 cross-section data and compared to the extrapolated or predicted critical mass for all the experiments. These experiments were performed at the Los Alamos National Laboratory using the Planet vertical lift critical assembly at the Los Alamos Critical Experiment Facility.
... This international compilation will be focused on high-priority needs. The format will be similar to the previously used United States format in that it will contain more specific information on the desired experimental features and parameters, as well as programmatic significance [3]. ...
New issues in criticality safety continue to emerge as spent fuel storage facilities reach the saturation point, fuel enrichments and burn-ups increase and new types of plutonium-carrying fuels are being developed. The new challenges related to the manipulation, transportation and storage of fuel demand further work to improve models predicting behaviour through new experiments, especially where there is a lack of data in the present databases. This article summarises the activities of the OECD/NEA working groups that co-ordinate and carry out work in the domain of criticality safety at the international level. Particular attention is devoted to establishing sound databases required in this area and to addressing issues of high relevance such as burn-up credit. The activities of working groups are aimed toward improving safety and identifying economic solutions to issues concerning the back-end of the fuel cycle.
The concrete is an important material to examine the calculation model employed in the evaluation of criticality safety for nuclear fuel cycle facility. Therefore, the data of the reflection and the isolation effects on reactivity were experimentally obtained using the Tank-type Critical Assembly (TCA). Critical experiments were performed for both the single and coupled cores where low-enriched-uranium fuel rods were arranged in a rectangular parallelpiped geometry. For the single core, the concrete slabs were positioned at one side of the core, whereas they were positioned between two cores for the coupled core to measure the reactivity interaction characteristics. Main parameters were selected as the thickness and the boron content of concrete, and the water-level worth method was applied to measure the reflection effect on the single core and the reactivity interaction effect from one core to the other core on the coupled system including symmetric and asymmetric arrangements. Some benchmark calculations were also executed using the SRAC code system to assess the accuracy of this code system. Moreover, the dependency of the reactivity interaction effect on the concrete thickness and the core geometry was examined on the basis of the Avery's two-point model.
The ratio of ²³⁸U captures to ²³⁵U fissions (C8/F5) was measured for four kinds of light-water moderated UO2 lattice cores; 1.42S, 1.00S, 0.75S and 0.56S. These cores have a central tight lattice test region surrounded by a normal lattice driver region. In the test region, the volume ratios of water-moderator to fuel are 1.420, 1.000, 0.750 and 0.564, respectively. Measurements were carried out using uranium metal foils set in a fuel rod at the center of the test region. The radial distributions of relative C8 and F5 rates in the fuel rod were also measured for the 1.42S and 0.56S cores. Calculations were performed by using the continuous energy Monte Carlo code MVP with the JENDL-3 library. The calculated C8/F5’s were larger than the experimental ones, and the discrepancy between the calculated in-fuel C8 distribution and the experimental one was observed. From the results of the sensitivity analysis, it is found that the decrease of the resonance ²³⁸U capture cross section and the increase of the resonance ²³⁶U fission cross-section reduced the discrepancy between the calculation and the experiment in the C8/F5 ratios and their in-fuel distributions, respectively.
For the criticality safety designs of the storage pools which store large number of spent fuels, neutron poison materials are useful for increasing capacity of the storage pool by minimiz-ing the spacing between the fuel assemblies. Since few experimental data exist for examining the validation of calculation methods which evaluate the nuclear criticality safety of fuel storage using the borated SUS (B-SUS) plates, critical experiments with B-SUS plates in the water region of single and coupled cores were performed at the Tank-type Critical Assembly (TCA). The systematic data on the reactivity effect of B-SUS plate as a function of the thickness, the boron content and the position of B-SUS plates in the water region were obtained using the critical water level method. The interacting effect between the two cores was also measured to clarify the nuclear characteristics of the coupled two core system which is the simplest system among multiple unit system. Criticality calculations with the combination of a Monte Carlo code, KENO-IV, and the MGCL-137 neutron cross section library in JACS were carried out, and these calculations reproduced the reactivity effect of B-SUS plate within ±0.6%ΔK/K.
Experimental and computational studies have been performed on the temperature coefficients of reactivity in light-water moderated and reflected UO2 cores with soluble poisons such as boron and gadolinium. Experiments were carried out using the Tank-type Critical Assembly (TCA) in Japan Atomic Energy Research Institute (JAERI). Temperature coefficients of the cores with soluble poisons were measured by changing the temperature of the moderator and reflector from the room temperature to about 60°C. The dependence of temperature coefficients on the core configuration and the concentration of soluble poison was studied with the water level worth method. Temperature coefficients were calculated with a diffusion code CITATION included in the SRAC code system and a perturbation code CIPER for comparison with the experimental data. It was found that the temperature coefficients are always negative in the experimental cores (the water to fuel volume ratio (Vm/Vf) of 1.83) containing boron as soluble poison. On the other hand, the temperature coefficients become positive in the cores with gadolinium due to the deviation of the gadolinium absorption cross section from the l/v law and the neutron spectral shift with the increase in temperature.
To determine the static k (effective neutron multiplication factor) ranging from the critical to an extremely subcritical states, the exponential experiments were performed using various sizes of light-water moderated and reflected low-enriched UO2 lattice cores. For comparison, the pulsed neutron source experiments were also carried out. In the manner of the Gozani's bracketing method applied to the pulsed source experiment, a formula to obtain k from the measured spatial-decay constant was derived on the basis of diffusion theory. Parameters in the formulas needed to obtain k from the respective experiments were evaluated by 4-group neutron diffusion calculations. The results of the exponential experiments agreed well with those of the pulsed source experiments, the 4-group diffusion calculations and the 137-group Monte Carlo calculations. Therefore, the present data-processing method developed for the exponential experiment was demonstrated to be valid. Besides, through the examination on the parameters used in the data processing, it was found that the dependence of parameter value upon k is weak in the exponential experiment compared with that in the pulsed source experiment. This indicates the superiority of the exponential experiment over the pulsed source experiment for the subcriticality determination of a wide range.
A study on a light-water moderated and reflected coupled-core was carried out to investigate the critical mass and neutron flux distribution as a function of the separation distance between the two cores. The critical mass reached the minimum value in a coupled-core with an approximately 1 cm wide water gap, and the peak value of neutron flux in the water gap region became the maximum when the separation distance is around 5 cm. These data were analyzed with the SRAC system. The calculated critical mass agreed well with those of experiments within an accuracy of approximately equals 1% . To calculate the neuron flux distribution, it was confirmed that attention should be paid to the number of energy groups and the method employed in the generation process of homogenized cross sections. It was found that the calculation can approximately reproduce the measured neutron flux distribution.
A criticality safety study on a light water moderated and reflected coupled core loaded with highly enriched uranium fuel was performed in the Kyoto University Critical Assembly. The critical mass and neutron flux distribution were measured systematically as a function of the separation distance between the two cores, varying the H/235U atomic ratio (i.e., the moderator-to-fuel volume ratio). These data were analyzed with the SRAC code system to assess the capability of diffusion theory to analyze the coupled-core system.
To establish a technique of on-site subcriticality determination suitable for the criticality safety management of nuclear fuel assembly, the applicability of the method proposed by Mihalczo was examined with the Tank-type Critical Assembly (TCA) of the Japan Atomic Energy Research Institute. In the Mihalczo method, cross power spectral densities and auto power spectral densities are evaluated from the output currents of an ionization chamber containing 252Cf neutron source and two neutron detectors. The principle of this method is that the spectral ratio formed by the power spectral densities mentioned can be related to the subcriticality by the help of a stochastic theory.Throughout our data processing, an improved formula taking account of the neutron extinction at a detection process was used. Up to the subcriticality of 15 dollars, the Mihalczo method agreed with the water-level worth method, which has been a standard method of reactivity determination at the TCA facility. The systems treated in the present report hold symmetry concerning the nuclear fuel configuration and the 252Cf chamber position. It was clarified that, contrary to Mihalczo's assertion, the factor converting the spectral ratio to a subcriticality depends on subcriticality itself.
This brochure is a basic source of information for prospective users of Sandia National Laboratories Radiation Facilities. It contains a brief description of the various major radiation sources, a summary of their output characteristics, and additional information useful to experimenters. Radiation source development and source upgrading is an ongoing program with new source configurations and modes of operation continually being devised to satisfy the ever-changing radiation requirements of the users. For most cases, the information presented here should allow a potential user to assess the applicability of a particular radiation facility to a proposed experiment and to permit some preirradiation calculations and planning.
Changes in the sign of the reactivity coefficient due to changes in the water density are very important in all water-moderated lattices. This sign is determined by the lattice parameters, such as the fuel enrichment and the lattice pitch. It is negative in undermoderated lattices. However, in special lattices, for example, in spent-fuel storage pools, this reactivity coefficient could be positive, even though one would predict a negative value from the lattice parameters. An example of this effect is presented, and the unexpected sign is explained.
In the nuclear criticality safety design of a nuclear fuel cycle facility, the geometric buckling of the fuel core is one of the most important quantities used in estimating criticality. When the material buckling value is known for a system consisting of fissile materials, it is possible to judge whether or not the system is subcritical by comparing the material buckling with the geometric buckling. It is widely known that the geometric buckling of a given core can be calculated by using a simple formula for some geometries, e.g., square, cylinder, slab, and sphere.
The experimental results of the geometrical buckling for typical regular polygons are described. Geometric buckling for three types of regular polygons has been measured in light-water-moderated UO2-H2O lattices in the tank-type critical assembly at the Japan Atomic Energy Research Institute. Based on the known critical buckling of a given experimental lattice and the measured critical water levels, the horizontal buckling has been evaluated for various sizes of regular hexagonal, square, and regular triangular cores. This method is based on the separability of geometric buckling into horizontal and vertical components. From the measured critical water levels of each core, it was found that the horizontal buckling of the effective core region is inversely proportional to the square of the radius of the circumscribed circle of the core. The geometric buckling B²g can therefore be expressed in the form of (aN/Rc)² using the geometric constant aN. The data for geometric buckling values on these geometries are available for the validation of calculation codes, and the empirical formula for geometric buckling obtained in this study can be applied to the basic criticality safety design of fuel cycle facilities.
A procedure is presented for the determination of geometric buckling for regular polygons. A new computation technique, the multiple reciprocity boundary element method (MRBEM), has been applied to solve the one-group neutron diffusion equation. The main difficulty in applying the ordinary boundary element method (BEM) to neutron diffusion problems has been the need to compute a domain integral, resulting from the fission source. The MRBEM has been developed for transforming this type of domain integral into an equivalent boundary integral. The basic idea of the MRBEM is to apply repeatedly the reciprocity theorem (Green’s second formula) using a sequence of higher order fundamental solutions. The MRBEM requires discretization of the boundary only rather than of the domain. This advantage is useful for extensive survey analyses of buckling for complex geometries.
The results of survey analyses have indicated that the general form of geometric buckling is B²g = (aN/Rc)², where Rc represents the radius of the circumscribed circle of the regular polygon under consideration. The geometric constant aN depends on the type of regular polygon and takes the value of π for a square and 2.405 for a circle, an extreme case that has an infinite number of sides. Values of aN for a triangle, pentagon, hexagon, and octagon have been calculated as 4.190, 2.821, 2.675, and 2.547, respectively.
Although the discussion is restricted to simple regular polygons, the proposed solution technique based on the MRBEM can easily be applied to many other complex geometries.
To get information about the neutron spectrum in low enriched UO2-H2O lattices, the spectral indices SI(U/Dy) and SI(U/U) were measured on the basis of the parallel irradiation technique, which basically irradiates activation foils both in a neutron field to be investigated and in a reference field of thermal neutrons. In the present study, a fuel pellet of UO2 was used for the measurement of activities caused by the neutron capture of U and the fission of U. Besides the technical details of the measurements, the origins of experimental errors are listed with the method how to eliminate them. The measurements were carried out in lattices of different fuel enrichment to demonstrate the capability of the present method, and the experimental results were compared with the calculated ones. It was found that the results of the present measurements are useful to assess the validity of the cell calculations.
A research program was initiated for the US Department of Energy (DOE) Sandia National Laboratory Transportation Systems Development Department in 1982 to provide benchmark type experimental criticality data in support of the design and safe operations of nuclear fuel transportation systems. The overall objective of the program is to identify and provide the experimental data needed to form a consistent, firm, and complete data base for verifying calculational models used in the criticality analyses of nuclear transport and related systems. A report, PNL-6205, issued in June 1988 (Bierman 1988) covered measurement results obtained from a series of experimental assemblies (TIC-1, 2, 3 and 4) involving neutron flux traps. The results obtained on a fifth experimental assembly (TIC-5), modeled after a calculational problem of the Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) Committee on the Safety of Nuclear Installations (CSNI) Working Group, are covered in this report. 10 refs., 10 figs., 7 tabs.
The modified conversion ratio is defined as the ratio of ²³⁸U captures to total fission. Gamma-ray spectrometry of irradiated fuel rods has been introduced to measure this quantity in two types of water-moderated low-enriched UO2 cores: the standard core, called the 1.42S core, and a tight-lattice core, called the 0.56S core. The water moderator-to-fuel volume ratios Vm/Vf of the cores are 1.420 and 0.564, respectively.
As no activation foil is used in this method, no corrections are needed for the neutron self-shielding and neutron flux depression that are caused by such a foil. Instead, the gamma-ray self-shielding effect due to the fuel rod must be corrected.
The modified conversion ratio is measured by this method are 0.457 for the 1.42S core and 0.724 for the 0.56S core. The errors in the experimental results are estimated to be∼3%. Computer analyses using the VIM continuous-energy Monte Carlo code with the JENDL-2 library show that the calculated value is ∼6% larger than the experimental one for the tight-lattice 0.56S core.
The Feynman-{alpha} experiments have been carried out using light-water-moderated and -reflected cores loaded with highly enriched uranium fuel at the Kyoto University critical assembly. An experimental technique using a multichannel scaler was developed to improve the accuracy of measurement and to shorten measuring time. Then, the {beta}{sub eff}/l values of single and coupled cores with different neutron spectra were measured to demonstrate the capability of the present technique for measuring the prompt neutron decay constant {alpha}. Moreover, the Feynman-{alpha} method was applied to measuring large subcriticalities. Through these experiments, it is found that the present technique greatly improves the accuracy of {alpha} measurement, and the one-point reactor approximation is applicable to a tightly coupled core. It is also found that the subcriticality down to approximately {minus} 35 $ can be measured by this method if the position of the neutron detector is chosen carefully, and the present Feynman-{alpha} method can be applied to a subcriticality monitoring system.
Results of recent fissile small sample worth experiments in three fast reactor diagnostic critical assemblies are presented. These experiments produced significant insight into heterogeneityrelated errors in standard calculational models of worth experiments in plate-type critical assemblies. Results of improved techniques for calculation of worth experiments are presented, and the mean ratios of calculated to experimental worths in the three new benchmark critical assemblies are shown to be in the range from 0.97 to 1.04. The implications of the improved understanding of small sample worth experiments with respect to previously reported critical experiments are discussed.
Biases and uncertainties of calculated reactor design quantities caused by errors and uncertainties of basic parameters, such as neutron cross sections, fission spectra parameters, and prompt and delayed neutron yields, are large, and in most cases, exceed reactor design requirements. Errors and uncertainties due to models and methods approximations contribute as well. An extensive data base, with presently /approximately/300 experimental integral values from 28 critical assemblies, has been assembled at Argonne National Laboratory in order to provide improvements and to investigate both sources of uncertainties. Generalized-least-squares fitting is being used. The available large data base permitted the investigation of the influence of specific input data, the constraints of the covariance information, the selection of parameters, and the reliability of the predictions. It is shown that reliable improvements of calculated quantities like enrichment, breeding ratio, sodium void, control rod worth, power distribution, and material worth can be made. Substantial reductions of the uncertainties of these quantities, which are caused by the uncertainties of the basic parameters, are obtained in most cases. The FFTF uranium-metal-core conversion is the first application of the present effort. 21 refs., 2 figs., 10 tabs.
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.
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.
The characterization data obtained to date are described for Approved Testing Material 104 (ATM-104), which is spent fuel from Assembly DO47 of the Calvert Cliffs Nuclear Power Plant (Unit 1), a pressurized-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-104 consists of 128 full-length irradiated fuel rods with rod-average burnups of about 42 MWd/kgM and expected fission gas release of about 1%. A variety of analyses were performed to investigate cladding characteristics, radionuclide inventory, and redistribution of fission products. Characterization data include (1) fabricated fuel design, irradiation history, and subsequent storage and handling history; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM) and electron probe microanalyses (EPMA); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding.
STACY and TRACY: Nuclear Criticality Facilities Under Construction
- I Kobayashi
I. Kobayashi et al., "STACY and TRACY: Nuclear Criticality Facilities Under Construction,"
Proceedings of International Conference of Nuclear Criticality Safety (ICNC'91), Oxford
Volume I, I-9 (1991).
Neutronic Design of Criticality Facilities (STACY and TRACY ) at JAERI for Criticality Safety Research
- Y Miyoshi
Y. Miyoshi et al., "Neutronic Design of Criticality Facilities (STACY and TRACY ) at JAERI
for Criticality Safety Research," Proceedings of 1988 Int. Reactor Physics Conf., Jackson Hole,
Vol. II, 493-504 (1988).
Design of Solution Criticality Facility of NUCEF
- I Izawa
I. Izawa et al., "Design of Solution Criticality Facility of NUCEF," Proceedings of International
Seminar on Nuclear Criticality Safety (ISCS'87),Tokyo, 38-43 (1987).
Critical Experiments on Low-enriched Uranyl Nitrate Solution of Cylindrical Core at STACY
- Y Miyoshi
Y. Miyoshi et al. "Critical Experiments on Low-enriched Uranyl Nitrate Solution of Cylindrical
Core at STACY," Proceedings of International Conference of Nuclear Criticality Safety
(ICNC'95), Albuquerque (1991).
Design of Transient Criticality Facility of NUCEF
- Takeshita
Takeshita et al., "Design of Transient Criticality Facility of NUCEF," Proceedings of
International Seminar on Nuclear Criticality Safety (ISCS'87), Tokyo, 44-49 (1987).
Determination of Static Reactivity by Exponential and Pulsed Source Experiments
- Suzaki
Suzaki, "Determination of Static Reactivity by Exponential and Pulsed Source Experiments,"
Proceedings of Int. Conf. on Nucl. Criticality Safety (ICNC'91), Oxford, VI-16 (1991).
Subcriticality Determination by a New Time-Domain Correlation Experiment with a 252 Cf Neutron Source
- Y Nishina
- Yamane
Nishina, Y. Yamane et al., "Subcriticality Determination by a New Time-Domain Correlation
Experiment with a 252 Cf Neutron Source," Proceedings of Int. Workshop on Subcritical
Reactivity Measurements, Albuquerque, (1985).
Estimation of Subcriticality by Neutron Source Multiplication Method
- T Sakurai
- Suzaki
Sakurai, T. Suzaki et al., "Estimation of Subcriticality by Neutron Source Multiplication
Method," JAERI-Research 95-002, (1995), in Japanese.
Measurements and Analyses on Reactivity Effects of Absorber Rods in a Light-water Moderated UO 2 Lattices
- Y Murakami
- Miyoshi
Murakami, Y. Miyoshi et al., "Measurements and Analyses on Reactivity Effects of Absorber
Rods in a Light-water Moderated UO 2 Lattices," JAERI-M 85-032 (1985), in Japanese.
Research on Reactor Physics Using the Tank-type Critical Assembly (TCA) -Focusing on the Method for Reactivity Determination Based on Buckling Measurements
- Suzaki
Suzaki, "Research on Reactor Physics Using the Tank-type Critical Assembly (TCA) -Focusing on the Method for Reactivity Determination Based on Buckling Measurements,"
KURRI-TR-305, 21 (1988), in Japanese.
20
Measurements and Calculations of Neutron Interaction Effects of a Two-coupled System in Water
- T Miyoshi
- Suzaki
Miyoshi, T. Suzaki et al., "Measurements and Calculations of Neutron Interaction Effects of a
Two-coupled System in Water," JAERI-M 90-112 (1990), in Japanese.
Measurements of Reactivity Effects of Neutron Absorber Materials Inside an Annular Core
- H Miyoshi
- Yanagisawa
Miyoshi, H. Yanagisawa et al., "Measurements of Reactivity Effects of Neutron Absorber
Materials Inside an Annular Core," JAERI-M 92-158 (1992), in Japanese.
Measurement of Reactivity Effect of Iron Plate Reflector in Light-water Moderated Low-enriched UO 2 Lattices
- T Murakami
- H Suzaki
- Hirose
Murakami, T. Suzaki, H. Hirose, "Measurement of Reactivity Effect of Iron Plate Reflector in
Light-water Moderated Low-enriched UO 2 Lattices," JAERI-M 83-100, (1983), in Japanese.
Evaluation of Temperature and Void Coefficients of Reactivity in Homogeneous Solution Fuel Systems
- Y Suzaki
- H Miyosh
- Hirose
Suzaki, Y. Miyosh, H. Hirose, "Evaluation of Temperature and Void Coefficients of Reactivity
in Homogeneous Solution Fuel Systems," Proceedings of Int. Sem. on Nuclear Criticality
Safety (ISCS'87), Tokyo, 383 (1987).
Measurements of Critical Masses of Non-uniform Fuel Rod Lattice Configurations
- T Yanagisawa
- K Suzaki
- Nitta
Yanagisawa, T. Suzaki, K. Nitta, "Measurements of Critical Masses of Non-uniform Fuel Rod
Lattice Configurations," ibid., 84 (1987).
Measurements and Calculations on Sloshing Effects in Solution Vessels Having a Free Surface
- H Miyoshi
- T Hirose
- Suzaki
Miyoshi, H. Hirose, T. Suzaki, "Measurements and Calculations on Sloshing Effects in
Solution Vessels Having a Free Surface," Proceedings of Int. Sem. on Nuclear Criticality
Safety (ISCS'87), Tokyo, 50 (1987).
Critical Sizes of Light-water Moderated UO 2 and PuO 2 -UO 2 Lattices
- I Tsuruta
- Kobayashi
Tsuruta, I. Kobayashi et al., "Critical Sizes of Light-water Moderated UO 2 and PuO 2 -UO 2
Lattices," JAERI 1254, (1978).
Computational Study on the Buckling-reactivity Conversion Factor in Light-water Moderated UO 2 Core
- Yamamoto
Yamamoto, "Computational Study on the Buckling-reactivity Conversion Factor in
Light-water Moderated UO 2 Core," JAERI-M93-170 (1993).
Dual and Multiple Reciprocity Formulations Applied to Fission Neutron Source Problems
- C A Itagaki
- Brebbia
Itagaki, C.A. Brebbia, "Dual and Multiple Reciprocity Formulations Applied to Fission
Neutron Source Problems," Proceedings of 14th Int. Conf. on BEM, Sevilia, Vol. l, 25 (1992).
Large Subcriticality Measurement by Pulsed Neutron Method
- A Yamane
- Yoshida
Yamane, A. Yoshida et al., "Large Subcriticality Measurement by Pulsed Neutron Method,"
Proceedings of Workshop on Subcriticality Reactivity Measurement, University of New
Mexico, Albuquerque, 128-147 (1985).
The Determination of Large Subcriticality by Pulsed Neutron Method for Nuclear Criticality Safety Study
- T Yamane
- Maki
Yamane, T. Maki et al., "The Determination of Large Subcriticality by Pulsed Neutron Method
for Nuclear Criticality Safety Study," Proceedings of Topical Meeting. on Reactor Physics and
Safety, Saratoga Springs, New York, September 17-19 1986, 1127.
Nuclear Criticality Safety Studies in the Kyoto University Critical Assembly, KUCA
- K Kanda
- Kobayashi
Kanda, K. Kobayashi et al., "Nuclear Criticality Safety Studies in the Kyoto University
Critical Assembly, KUCA," Proceedings of Int. Seminar on Nucl. Criticality Safety (ISCS'87)
October 19-23, Tokyo, 37 (1987).
Study on Coupled-core in the Kyoto University Critical Assembly
- S Misawa
- K Shiroya
- Kanda
Misawa, S. Shiroya, K. Kanda, "Study on Coupled-core in the Kyoto University Critical
Assembly," ibid. (ISCS'87), 58-64 (1987).
Measurement of Large Subcriticality by Pulsed Neutron Method for Nuclear Criticality Safety Study
- C Kobayashi
- Ichihara
Kobayashi, C. Ichihara et al., "Measurement of Large Subcriticality by Pulsed Neutron Method
for Nuclear Criticality Safety Study," ibid. (ISCS'87), 301-306 (1987).
Measurement of Reactivity Effect Caused by Non-uniform Fuel Distribution
- Y Yamane
- Hirano
Yamane, Y. Hirano et al., "Measurement of Reactivity Effect Caused by Non-uniform Fuel
Distribution," Intl. Conf. on Criticality Safety, September 9-13 1991, Oxford, UK (1991).
Nuclear Data Qualification Through French LWR Integral Experiments
- Santamarina
Santamarina, "Nuclear Data Qualification Through French LWR Integral Experiments,"
Proc. Int. Conf. on Nuclear Data, 1, 509. Santa Fe (NM), USA, 1985.
CEA-86" Multigroup Cross-section Library and Its Integral Qualification
- H Santamarina
- Tellier
Santamarina and H. Tellier, "The French "CEA-86" Multigroup Cross-section Library and Its
Integral Qualification," Proc. Int. Conf. on Nuclear Data, Mito, Japan, 30 May -3 June, 1988.
The CAMELEON Experiment
- A Martin-Deidier
- Santamarina
Martin-Deidier, A. Santamarina et al., "The CAMELEON Experiment," Transactions ANS, 46,
755 (1984).