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NEA-1687 IRPHE/B&W-SS-LATTICE

IRPHE/B&W-SS-LATTICE, Spectral Shift Reactor Lattice Experiments

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1. NAME OR DESIGNATION OF PROGRAM:  IRPHE/B&W-SS-LATTICE.
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2. COMPUTERS
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Program name Package id Status Status date
IRPHE/B&W-SS-LATTICE NEA-1687/04 Tested 12-DEC-2003

Machines used:

Package ID Orig. computer Test computer
NEA-1687/04 Many Computers PC bi-Pentium III 1GHz
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3. DESCRIPTION OF PROGRAM OR FUNCTION

B&W has performed and analysed a series of physics experiments basically concerned with the technology of heterogeneous reactors moderated and cooled by a variable mixture of heavy and light water. A reactor so moderated is termed Spectral Shift Control Reactor (SSCR). In the practical application of this concept, the moderator mixture is rich in heavy water at the beginning of core life, so a relatively large fraction of the neutrons are epithermal and are absorbed in the fertile material. As fuel is consumed, the moderator is diluted with light water. In this way the neutron spectrum is shifted, thereby increasing the proportion of thermal neutrons and the reactivity of the system.
  
The general objective of the SSCR Basic Physics Program was to study the nuclear properties of rod lattices moderated by D2O-H2O mixtures. The volume ratio of moderator to non-moderator in all lattices was approximately 1.0, and the fuel was either 4%-enriched UO2 clad in stainless steel or 93%-enriched UO2-ThO2 (NTh/N25 = 15) pellets clad in aluminum. The D2O concentration in the moderator ranged from zero to about 90 mole %. The experimental program includes critical experiments with both types of fuel, exponential experiments at room temperature with both types of fuel, exponential experiments at elevated temperatures with the 4%-enriched UO2 fuel, and neutron age measurements in ThO2 lattices. The theoretical program included the development of calculation methods applicable to these systems, and the analysis and correlation of the experimental data.
  
A first report provides the results of critical experiments performed under the Spectral Shift Control Reactor Basic Physics Program are documented here. Nine major critical assemblies of rod lattices were studied in moderator mixtures of light and heavy water ranging from zero to 81. 2 mole % D2O. In some assemblies, the moderator was poisoned with boric acid. The non-moderator-to-moderator volume ratio in all lattices was approximately-1.0. The fuel in most lattices was 4%-enriched UO2 swaged in stainless steel, although two experiments were performed with 93%-enriched UO2- ThO2 pellets in aluminum tubes,. One assembly was zone-loaded radially and contained both types of fuel.
  
The critical mass, D2O concentration, boron concentration, buckling, thermal disadvantage factor, and cadmium ratio of U-235 were measured in each assembly, In most assemblies, the cadmium ratio of U-238 or Th-232 was measured, and in five assemblies, the epithermal neutron spectrum was derived from the measurements taken of the resonance activity of detector foils In special experiments at high D2O concentrations, the perturbations by moderator gaps and control blades were studied, and the reflector savings versus reflector thickness was measured. The flux distribution in the zone-loaded assembly was also mapped.
  
A second report documents experimental results and theoretical interpretation of a series of twenty uniform lattice critical experiments in which the neutron spectrum is varied over a fairly broad range. Two types of fuel rods were studied: 4.02%-enriched UO2 in stainless steel tubes and 2.46% enriched UO2 in aluminum tubes, Lattice nonmoderator-to-moderator volume ratios ranged from 0.65 to 1. 2. The moderators were mixtures of light and heavy water ranging in composition from zero to 77% D2O, with and without boric acid. Measurements include critical size and composition, r/h, buckling and reflector savings, thermal disadvantage factor, and cadmium ratios of U-235 and U-238   Theoretical methods used to analyze the data are given, and results are compared.
  
A third report addresses  issues that bear on the problems associated with plutonium utilisation in commercial reactors is being integrated into the program. The results of research conducted by Babcock & Wilcox is documented. It concerns specifically plutonium lattice experiments in uniform test lattices of UO2 1.5% PuO2 fuel. The system used was tailored to provide a critical assembly with a small central test region having a nearly flat flux and asymptotic spectrum. The test region was incrementally poisoned with boric acid until a void substitution yields a null reactivity change. This gives a precise measure of the boron concentration required to reduce to unity the infinite medium multiplication factor of the test lattice. Comparison of this number with the corresponding analytical prediction is a sensitive and direct check on the adequacy of the theoretical procedures. Various other measurements are also taken on the poisoned and unpoisoned test lattice for other tests of the analytical methods and to facilitate the deduction of kinf for the unpoisoned lattice. This report describes the facility, the equipment, and the techniques of measurement and presents the results of the measurements.
  
A fourth report concerns Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel. Close-packed storage of LWR fuel assemblies is needed in order to expand the capacity of existing underwater storage pools. This increased capacity is required to accommodate the large volume of spent fuel produced by prolonged onsite storage. To provide benchmark criticality data in support of this effort, 20 critical assemblies were constructed that simulated a variety of close-packed LWR fuel storage configurations. Criticality calculations using the Monte Carlo KENO-IV code were performed to provide an analytical basis for comparison with the experimental data. Each critical configuration is documented in sufficient detail to permit the use of these data in validating calculational methods according to ANSI Standard N16.9-1975.
  
The fifth report concerns critical experiments supporting underwater storage of tightly packed configurations of spent fuel pins. Critical experiments have been performed with low-enriched UO2 arrays simulating underwater pin storage of spent pressurized water reactor fuel. "Pin storage" refers to a storage concept in which fuel assemblies are dismantled and the individual fuel pins from several assemblies are tightly packed into specially designed cannisters. Each critical configuration is sufficiently described and documented to permit the use of these data for validating criticality calculational methods according to ANSI Standard N16.9-1975. The reactivity of each benchmark core was calculated using the AMPX-KEN0 IV package. The results of these analyses are also presented.
  
The sixth report entitled "PHYSICS VERIFICATION PROGRAM", covers principally a series of experiments to measure the effect of lattice heterogeneities. Part of this was revisited in an ANS benchmark study in 1996, a specification and the results from participants are all included as separate directories in this package. This benchmark has been proposed for further analysis within the Reactor Physics Division of ANS in November 2002. The revised specification is provided.
  
The seventh report concerns Development and Demonstration o f an Advanced Extended-Burnup Fuel- Assembly Design Incorporating Urania-Gadolinia. Comparisons of the thermal properties of UO2-Gd2O3 specimens containing 2.98, 5.66, and 8.50 w t % Gd2O3 with UO2 specimens showed that thermal conductivity is the only thermal parameter Significantly affected by the addition of Gd2O3. The milling steps used to prepare UO2-Gd2O3 powder result in a powder that is more active than standard UO2 powder. As a result, UO2-Gd2O3 fuel has shown more variability than UO2 fuel in as-sintered theoretical density and densification behavior. However, a porefoming material, added to the UO2-Gd2O3 powder mixture before sintering, can be used to achieve the desired density.
  
Measured results from critical experiments were compared with predicted data and confirmed t h e accuracy o f the standard two-group diffusion theory model for predicting global and discrete UO2-Gd2O3 effects when cross-section input is appropriately adjusted. The preliminary first two fuel cycles for lead test assemblies of the advanced design were developed. Irradiation of the lead test assemblies will be used irradiation assemblies is scheduled to begin in 1983 in Duke Power Company's Oconee Unit 1. An inter-calibrated movable in-core detector system to monitor the performance of the test assemblies during irradiation.
  
The eighth report concerns "Urania-Gadolinia: Nuclear Model Development and Critical Experiment Benchmark". The purpose was development and verification within an extended-burnup program for pressurised water reactors of  advanced fuel assembly design. This advanced fuel assembly uses a UO2-Gd2O3 burnable absorber fuel mixture along with other state-of-the-art fuel performance and uranium utilisation-enhancing design features including annular fuel pellets, annealed guide tubes, Zircaloy intermediate grids, and removable upper end fittings. As part of this program, a nuclear model for calculating the behavior of UO2-Gd2O3 was developed, and critical experiments were conducted to provide beginning-of- life data for benchmarking the model. The report describes the nuclear model, the critical experiments, and the comparison of data between the model and the experiments. The comparison confirmed the accuracy of the standard two-group, diffusion theory model for predicting global and discrete UO2-Gd2O3 effects provided the cross sections used in the model were generated with sufficiently detailed and sophisticated methods.
  
The ninth report concerns the Characterization and Irradiation Program: Extended-Burnup Gadolinia Lead Test Assembly (Mark GdB). The goal of the program was to extend the burnup of pressurized water reactor fuel assemblies to 50,000 MWd/mtU batch average burnup. To achieve this goal, five lead test assemblies have been designed, manufactured, characterized, and inserted for irradiation in Oconee Unit 1 cycle 8. One lead test assembly received 13,989 MWd/mtU burnup by its discharge at the end of cycle 8. Three of the four remaining lead test assemblies received approximately 45,000 MWd/mtU burnup by their discharge at the end of cycle 100. The fourth lead test assembly received approximately 58,000 MWd/mtU before being discharged at the end of cycle 11.

The tenth report concerns the Hot Cell Examination of Gadolinia Lead Test Assembly Rods After One Cycle of Irradiation as described in the eighth and ninth report. To track the performance characteristics of the advanced assemblies, non-destructive examinations at poolside were conducted after each cycle of irradiation .Nondestructive and destructive examinations were conducted in the hot cell after the first cycle of irradiation. This report contains the data collected during the hot cell examination of 17 fuel rods taken from the discharged gadolinia lead test assembly after one cycle of irradiation. Where appropriate, those data are compared with the data from the pre-irradiation characterization and the data base for the standard Mark B fuel assemblies.
  
The eleventh report covering April 1986 through March 1987, combines the progress report for the program entitled Development of an Advanced Extended Burnup Fuel Assembly Design Incorporating Urania-Gadolinia, and the final progress report for the program entitled Qualification of the B&W Mark B Fuel Assembly for High Burnup.  Under the latter program, standard Mark B fuel assemblies were irradiated and examined after three, four, and five cycles of operation to provide fuel performance data at extended burnup. During this reporting period, four gadolinia lead test assemblies (LTAs) . were examined at poolside after their second cycle of irradiation. These four assemblies were inserted for a third cycle of irradiation, which has reached 266 effective full-power days. Also, comparisons of data for cladding waterside oxide thickness of standard Mark B fuel rods with predictions from a simplified oxide buildup model were performed. Results indicate that this model adequately predicts the oxide thickness in the maximum oxide region of the fuel rods. However, the model's sensitivity to temperature suggests that improvements are necessary for generic application.
  
The twelfth report describes five lead test assemblies designed, manufactured, characterized, and inserted for irradiation in Oconee Unit 1 cycle 8. One lead test assembly received a burnup of 13,900 MWd/mtU by its discharge at the end of cycle 8; the remaining four received a burnup of 16,000 MWd/mtU. Of the four remaining lead test assemblies, three received burnups of approximately 45,000 MWd/mtU by their discharge at the end of cycle 10. The fourth lead test assembly received approximately 58,300 MWd/mtU when discharged at the end of cycle 11. To track the performance characteristics of the advanced assemblies, non-destructive examinations at poolside were conducted. This report contains the data collected during the poolside examination of the lead test assemblies after three cycles of irradiation.
  
Core characteristics
  
Core    Number of    Fuel enrichment
        fuel Rods    wt% U235
I        484-U       4.02
II      4904-U       4.02
III     5284-U       4.02
III-21  3544-U
III-22  5284-U       4.02
IV      2252-U       4.02
V       5284-U       4.02
VI      5284-U       4.02
VII-A   2704-Th
VII-B   5212-U       4.02
VIII    2188-Th
IX       952-U       4.02
X        608-U       4.02
XI      5320-U       4.02
XII     1390-U       4.02
XIII     596-U       2.46
XIV     2852-U       2.46
XV      1140-U       2.46
XVI     5124-U       2.46
XVII     872-U       2.46
XVIII   5137-U       2.46
XIX     5137-U       2.46
XX      5137-U       2.46
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9. STATUS
Package ID Status date Status
NEA-1687/04 12-DEC-2003 Screened
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12. PROGRAMMING LANGUAGE(S) USED
No specified programming language
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15. NAME AND ESTABLISHMENT OF AUTHORS

   Babcock & Wilcox Co.
   P.O. Box 10935
   LYNCHBURG, VA 24506-0935
   U.S.A.
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16. MATERIAL AVAILABLE
NEA-1687/04
Babcock & Wilcox Lattice Experiments
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17. CATEGORIES
  • Y. Integral Experiments Data, Databases, Benchmarks

Keywords: criticality, experimental data, lattice, reactor cores.