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Hunter
Journal of Nuclear Science and Technology, 37(10), p.855 - 869, 2000/00
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Hunter
JNC TN9400 99-049, 74 Pages, 1999/04
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%
k/k higher than the measured value, Na void worth C/E values were 
1.06; these results were considered to be reasonable. (Discrepancies in certain Na void values were probably due to experimental causes, though the efect should be investigated in any future experiments.) several sample worth values were small compared with calculational uncertaint
Hunter
PNC TN9410 98-015, 81 Pages, 1998/02
The study was carried out within the framework of the PNC-CEA collaboration agreement. Data were provided, by CEA, for an experimental loading of a start-up core in Super-Phenix. This data was used at PNC to produce core flux snapshot calculations. CEA undertook a comparison of the PNC results with the equivalent calculations carried out by CEA, and also with experimental measurements from SPX. The resu1ts revealed a systematic radial flux tilt between the calculations and the reactor measurements, with the PNC tilts only 30-401 of those from CEA. CEA carried out an analysis of the component causes of the radial tilt. It was concluded that a major cause of radia1 tilt differences between the PNC and CEA calculations lay in the nuclear datasets used: JENDL-3.2 and CARNAVAL IV. For the final stage of the study, PNC undertook a sensitivity analysis, to examine the detailed differences between the two sets of nuclear data. The PNC flux calculations modelled SPX in both 2D (RZ) and 3D (hex-Z) geometries, using the diffusion programs CITATION and MOSES. The sensitivity analysis of the differences between the JENDL-3.2 and CARNAVAL IV nuclear datasets used the SAGEP calculational route. Both datasets were condensed to a single, non-standard, set of energy group boundaries. There were some incompatibilities in the cross-section formats of the two datasets. The sensitivity analysis showed that a relatively small number of nuclear data items contributed the bulk of the radial tilt difference between calculations with JENDL-3.2 and with CARNAVAL IV. A direct comparison between JENDL-3.2 and CARNAVAL IV data revealed the following. The Nu values showed little difference (<5|%). The only large fission cross-section differences were at low energy (<30% otherwise, with <10% typical). Although down-scattering reactions showed some large fractional differences, absolute differences were negligible compared with in-group scattering; for in-group scattering fractional ...
Hunter
PNC TN9460 98-001, 156 Pages, 1998/01
This document provides a description of a calculational route, used in the Reactor Physics Research Section for sensitivity studies and initial design optimization calculations for fast reactor cores. The main purpose in producing this document was to provide a description of and user guides to the calculational methods, in English, as an aid to any future user of the calculational route who is (like the author) handicapped by a lack of literacy in Japanese. The document also provides for all users a compilation of information on the various parts of the calculational route, all in a single reference. In using the calculational route (to model Pu burning reactors) the author identified a number of areas where an improvement in the modelling of the standard calculational route was warranted - the document includes a description of these changes. The calculational route makes use of several different computer programs. SLAROM calculates nuclear data from compositions, using either homogeneous or heterogeneous models. CITATION and MOSES do reactor burn-up and/or flux diffusion calculations; CITATION is used for 2D (RZ) calculations, whilst MOSES models 3D (hex-Z) geometry. PENCIL and CITDENS are essentially specialized versions of CITATION (PENCIL includes data preparation and other functions). MASSN calculates fuel cycle mass balances. PERKY performs perturbation and associated calculations, both 1'st order and exact perturbations. JOINT and RZOUT3 provide various dataset interface functions, including energy group condensation. Briefer descriptions of the calculational route are given, followed by a more detailed step-by-step approach to the calculations. This latter includes examples of all JCL and data files, and a description of all the data that a user may have to employ. The document does not give a complete description of the component programs: where options and/or data are not used in any of the calculations they have generally been ignored; ...
Hunter
PNC TN9410 97-057, 106 Pages, 1997/05
This study was based on a 'pancake' type fast reactor core design of 600 MW(e), which had been optimized for Pu burning with a feed Pu vector appropriate to once-through irradiation of MOX fuel in a PWR. The purpose of the study was to investigate the effects of varying the Pu vector, examining various methods of offsetting the effects of such a change, and finally to produce fuel cycles optimized for the different qualities of Pu vector within the same basic design. In addition to the reference (once-through) Pu vector, two extreme Pu vectors were examined: high quality Pu from military stockpiles; low quality Pu corresponding to the equilibrium point of multiple recycling in a Pu burning fast reactor. Variations in Pu quality were overcome by changing the fuel inventory - replacing some of the fuel by diluent material, and altering the fuel pin size. Using absorber material (
B
C) as diluent improves the rod worth shutdown margin but degrades the Na void and Doppler safety parameters, a non-absorber diluent has the opposite effects, so a mix of the 2 material types was used to optimize the core characteristics. Of the non-absorber diluent materials examined, ZrH gave significantly better performance than all others; 
BC was the second choice for non-absorber diluent, because of its compatibility with BC absorber. It was not possible to accommodate the lower quality (multi-recycled) Pu vector without a significant increase in the fuel pin volume. It was not generally possible, especially with the increased fuel pin size, to achieve positive rod worth shutdown margins - this was overcome by increasing the number of control rods. For the higher quality Pu vectors to maintain ratings within limits, it was necessary to adopt hollow fuel pellets, or else to use the diluent material as an inert matrix in the fuel pellets. It proved possible to accommodate both extremes of Pu vector within a single basic design, maintaining ...
Hunter
Proceedings of International Conference on the Physics of Reactors (PHYSOR '96), 0 Pages, 1996/09
None
Hunter, S.
PNC TN9410 96-011, 107 Pages, 1996/01
This study used a "pancake" type fast reactor core design of 600 MW(e) which had been optimised for Pu burning with a feed Pu vector appropriate to once-through irradiation of MOX fuel in a PWR. The purpose of the study was to investigate the feasibility of using different qualities of Pu vector within the same basic core design. In addition to the reference (once-through) Pu vector, two extreme Pu vectors were examined: high quality Pu from military stockpiles; low quality Pu corresponding to the equilibrium point of multiple recycling in a Pu burning fast reactor. The calculations used a 2D R-Z model. The cell lattice code SLAROM, the diffusion & burnup code CITATION and the perturbation code PERKY were used. Cross-section data was taken from JENDL-3.2. The key results for the reference Pu vector case were a reactivity loss per cycle of 4.304% and the main safety parameters, a Na void worth of 1.631% and a Doppler constant of -0.00573 (a ratio of -285). The reactivity loss and void-to-Doppler ratio were used as targets for the calculations with the other Pu vectors. For the high quality Pu vector it was necessary to introduce an amount of diluent material into the core, to allow criticality criteria to be met. Several different classes of diluent were examined: increased fuel pin voidage, materials transparent to neutrons, moderators and neutron absorber. Compared with the initial "reference" Pu vector, with absorber as diluent there was an improvement in the reactivity loss per cycle (down to 3.4%) but a degradation in the main safety parameters with the void : Doppler ratio around -2000. With any of the other materials used as diluent the situation was reversed, with the reactivity loss increased to around 6% but Na void and Doppler improved to a ratio around -20. A mixture of absorber with another diluent material was shown to balance the above effects. A diluent of ZrH
together with BC was Capable of producing an ...
; Sugino, Kazuteru; Hunter
NEA Workshop on the Physics and Fuel Peactor-Based Pluto-nium Disposition, ,
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Hunter; Tikhom, A. V.*; Semen, M. Yu.*; Rimpault, G.*
Proceedings of International Congress ENS '98 and World Exhibition, ,
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