ML13002A015
| ML13002A015 | |
| Person / Time | |
|---|---|
| Site: | University of Texas at Austin |
| Issue date: | 12/19/2012 |
| From: | Whaley P University of Texas at Austin |
| To: | Lising A Document Control Desk, Division of Policy and Rulemaking |
| References | |
| TAC ME7694 | |
| Download: ML13002A015 (51) | |
Text
DcLparcrnenc of Mechanical Engineering THE UNIVERSITY OF TEXAS AT AUSTIN Nuclear Engineering T7acbing Laboratory. Austin. Texas 78758 s
/512-232-5370 FAX 512-471-4589-htep:/Iwww.me.utexas.edul-netl/
December 19, 2012 ATTN: Document Control Desk, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 Allan Jason Lising Project Manager Division of Policy and Rulemaking Research and Test Reactors Licensing Branch
SUBJECT:
Docket No. 50-602, Request for Renewal of Facility Operating License R-129 REF: UNIVERSITY OF TEXAS AT AUSTIN - REQUEST FOR ADDITIONAL INFORMATION REGARDING THE LICENSE RENEWAL REQUEST FOR THE NUCLEAR ENGINEERING TEACHING LABORATORY TRIGA MARK II NUCLEAR RESEARCH REACTOR (TAC NO. ME7694)
Sir:
Attached are Report on Neutronic Analysis for the UT TRIGA Reactor and Historical UT TRIGA Core Data, which address some of the items in the referenced Request for Additional Information, including:
RAI 7: The guidance in NUREG-1537 Section 4.5, "Nuclear Design," requests that the licensee provide a detailed description of analytical methods used in the nuclear design, including computer codes used to characterize technical parameters pertaining to the reactor. UT SAR Section 4.5 states that the "characteristics and operating parameters of this reactor have been calculated and extrapolated using experience and data obtained from existing TRIGA reactors as bench marks in evaluating the calculated data." Please provide comprehensive analysis of UT TRIGA behavior. Please describe the methods used for steady state neutronic (steady-state and kinetics) and thermal-hydraulic analysis and include comparisons with UT TRIGA measurements that demonstrate that those methods are appropriate to analyze the limits imposed by the UT TRIGA TS.
RESPONSE
RAI 7 is partially addressed in Attachment 1, specifically the neutronic portion of the RAI, by Sections 2 (describing the SCALE computer code suite) and 4 (modeling the UT TRIGA reactor in SCALE) of the attachment. Thermo hydraulic modeling is not addressed in this response.
Validation of methodology and modeling is provided in Section 6.
RAI 8
The guidance in NUREG-1537 Section 4.5.1, "Normal Operating Conditions," requests that the licensee define the limiting core configuration (LCC) which defines the highest power densities and temperatures achievable.
RAI 8.1: UT SAR Section 4.5.1 discusses an "operational core of 85 fuel elements, 3 fuel followed control rods, and one air followed control rod is to be arranged in 5 rings." The NRC staff notes that elsewhere in the UT SAR, references are made to core configurations of 90 (page 4-36), 116 (page 4-55), 109 and 114 (page 4-57), 81 (page 13-7), 83 (page 13-2), 85 (pages 13-26 and 13-29), and 100 (page 13-33) fuel elements. For the UT TRIGA licensed power of 1,100 kW, please identify the LCC. Please provide schematic drawings showing the location of fuel elements, control rods, and other components installed in the lettered-and-numbered lattice positions.
For fuel elements provide a cross reference to fuel element serial numbers and their accumulated burnup. Please provide all technical parameters and conclusions supplied for normal operation, accident analysis, and dose estimates using the LCC.
RESPONSE
The limiting core configuration (LCC) is identified in Section 5 of the Attachment 1.
Schematics showing the location of core components are provided in Figs. 1-6, and 15 of. The LCC analysis assumes fresh fuel, and does not assume specific elements or burnup for initial operation. Burnup is discussed for the 1992 core in Sections 1.3 and 6.7, and illustrated on Fig. 3.
Technical parameters for normal operation, accident analysis, and dose estimates are provided in Section 5 of Attachment 1. The limiting core configuration is established, and the parameters are extended from the minimum number of fresh fuel elements that comprise the limiting core configuration through a fully-fueled core with burnup at end of core life.
Fuel element serial numbers, in-core locations, and accumulated burnup values are provided in.
RAI 8.2: Please provide analyses that quantify the effects of fuel burnup, plutonium buildup, and the effect of fission products on the UT TRIGA LCC.
RESPONSE
Effects of fuel burnup are discussed in Sections 5.2.1 and 5.2.3 of Attachment 1.
RAI 8.3: Please provide the technical parameters including analysis of "reactor kinetic behavior, basis reactor criticality, control rod worth, definition of the limiting core configuration (LCC), [etc.]"
(NUREG-1537, Section 4.5.1). State whether the comparison of calculated and measured values demonstrates acceptable model development.
RESPONSE
Please contact me by phone at 512-232-5373 or email whaley@mail.utexas.edu if you require additional information or there is a problem with this submittal.
Thank you, P. M. Whaley Associate Director Nuclear Engineering Teaching Laboratory The University of Texas at Austin I declare under penalty of perjury that the foregoing is true and correct.
Executed on December 12, 2012 Steven R. Biegalski NETL Director ATT:
The limiting core configuration is identified in Section 5 of Attachment 1, with considerations of critical mass in 5.1.1 and 5.1.2, kinetics parameters in Section 5.21, control rod worths in 5.2.6.
Validation of methodology and modeling is provided in Section 6.
RAI 9: The guidance in NUREG-1537 Section 4.5, "Nuclear Design," requests that the licensee provide a detailed description of analytical methods used in the nuclear design, including computer codes used to characterize technical parameters pertaining to the reactor. UT SAR Section 4.5 states that the "characteristics and operating parameters of this reactor have been calculated and extrapolated using experience and data obtained from existing TRIGA reactors as bench marks in evaluating the calculated data." Please provide comprehensive analysis of UT TRIGA behavior. Please describe the methods used for steady state neutronic (steady-state and kinetics) and thermal-hydraulic analysis and include comparisons with UT TRIGA measurements that demonstrate that those methods are appropriate to analyze the limits imposed by the UT TRIGA TS.
RESPONSE
RAI 7 is partially addressed by Attachment 1, specifically the neutronic portion of the RAI by Sections 2 (describing the SCALE computer code suite) and 4 (modeling the UT TRIGA reactor in SCALE) of the attachment. Thermal-hydraulic modeling is not addressed in this response.
Validation of methodology and modeling is provided in Section 6.
RAI 10: UT SAR Table 4.21, "Limiting Core reactivity," displays Reference and Current control rod worths. In Table 4.14, please explain the origin of the values listed under the "Reference" column. Given the difference between the "Reference" and "Current" values of excess reactivity and shutdown margin, which values are being used in the UT TRIGA TS.
RESPONSE
The Table will be deleted, with reference values used in model validation (Section 6 of ) and reactivity values for the LCC with burnup provided in Section 5.2.6.
RAI 12: UT SAR Section 4.5.4, Subsection B provides Figure 4.22 for the power within a fuel element.
The NRC staff notes that the power distribution in the figure continues to the center of the fuel element indicating that this curve is not applicable to stainless steel-clad fuel that has a zirc rod in the center. Please confirm and revise accordingly.
RESPONSE
Updated power distribution analysis is reported in in Section 5.2.2 of Attachment 1.
We respectfully request an additional 90 days to complete response to the remaining items.
ATTACHMENT 1:
REPORT ON NEUTRONIC ANALYSIS FOR THE UT TRIGA REACTOR The UT TRIGA critical mass of the original UT TRIGA reactor core configuration is compared with the critical mass required in prototypical cores. The computer codes used in this analysis is described. The geometry of the UT TRIGA core is identified. The representation of the core geometry and materials in modeling within the program is described. The results of calculations using the model to characterize the UT TRIGA reactor are summarized. Finally, evidence demonstrating validity of the model in characterizing the UT TRIGA reactor is provided.
The UT TRIGA reactor core uses a triangular pitch, composing a hexagonal geometry as shown in Fig. 1.
Core positions are indexed as rings (A through G), with index numbers increasing for sequential positions. Neutronic analysis was performed for the UT TRIGA reactor configured with three standard fuel follower control rods located in positions C01, C07, D06, and D14. Position A01 (central thimble) does not contain a standard fuel element. Positions G32 and G34 are reserved for a neutron source and an in-core pneumatic terminal. The A and B rings are within a removable assembly that allows insertion of large experiments; because of the associated reactivity deficit, removing the fuel in these positions severely limits potential operation. Fuel element positions E03, E04, F03, F04, F05, G04 and G05 can similarly be removed with full power operations possible. Two smaller removable "3-element" assemblies are located at the D17/E22/E23 and E11/F12/F14 positions.
626 627 G28 O29 G30 G12 all 010 609 Go0 Figure 1, UT TRIGA Reactor Core 1.0 COMPARISON WITH HISTORICAL CRITICAL MASS DATA 1.1 CRITICAL MASS OF HISTORICAL GENERAL ATOMICS TRIGA REACTORS The cylindrical GA TRIGA Mark I reactor achieved critical condition with 54 standard fuel elements (8.5%
weight uranium, 20% enriched, 1.94 kg 23SU) and four water filled control rod positions. Core positions not occupied by fuel elements used graphite "dummy" rods for reflection; removing the graphite rods increased critical mass about 25%. The GA Advanced TRIGA Prototype reactor achieved critical
condition with three fuel follower control rods, 1 transient rod, and 75 standard fuel elements (2.7 kg 235U). The GA TRIGA Mark III reactor achieved critical condition with four fuel follower control rods (including a fuel element in the "A" ring) and 56 standard fuel elements (2.24 kg 235U) 1.2 CRITICAL MASS OF THE 1992 UT TRIGA REACTOR Initial criticality for the UT TRIGA reactor at the Nuclear Engineering Teaching Laboratory (NETL) was accomplished on 02/13/1992. Criticality was attained with 3 fuel follower control rods (fully withdrawn, i.e., fuel fully inserted) and 56 fuel rods (including two instrumented fuel elements). Total mass of 23SU was 2.12 kg in the standard fuel elements and 94.46 g in the three fuel followers.
The UT TRIGA reactor was configured with both water voids and graphite rods in non-fueled positions.
With the exception of fuel follower elements (i.e., fuel follower control rods), the 1992 University of Texas (UT) TRIGA reactor core was composed of fuel elements with power history at the previous UT TRIGA (located on main campus in Taylor Hall) and/or a General Atomics facility. These lightly-burned fuel elements decayed approximately 1 year prior to use at the current location. The new UT TRIGA core included three new fuel follower elements with fresh fuel. Critical fuel loading is displayed in Fig. 2, with labels:
CT for the central thimble (water void)
SFE for standard fuel rods TC for instrumented fuel elements GR for graphite rods S for the neutron source WV for water voids 52, S1 and RR for the fuel follower control rods (Shim 1 and 2, Regulating Rod)
TR for the transient rod Although the pitch is hexagonal, positions are considered to be in "rings." The B ring is shaded green in Fig. 2, C ring light blue-green, D ring rose, E ring light brown, F ring light blue, and the G ring yellow.
A direct comparison between critical masses of the initial 1992 UT TRIGA core and the historical GA TRIGA cores is complicated by (1) differences in reflection (graphite rod and water void configurations),
(2) previous power history for the standard fuel elements, and (3) fundamental difference in core geometry. Nevertheless, the 2.2 kg 235U in the UT TRIGA compares well with the 1.94 kg (approximately 2.4 kg water moderated) of fresh elements required in the original GA TRIGA reactor.
WV WV WV WV WV W
V WV WV WV WV WV Figure 2, Initial UT TRIGA Core 1.3 OPERATIONAL LOADING OF THE 1992 UT TRIGA REACTOR Fuel was loaded in the UT TRIA reactor to support operation at 1.1 MW on 03/16/1992. The core contained 84 standard and 3 fuel follower elements with 3.35 kg 235U. Operational fuel loading is displayed in Fig. 3, using the same labeling as in Fig. 2.
GR GR GR GR GR Figure 3, Operational 1992 Core Power history for the standard fuel elements in this configuration included 7.91 MWD generated at Taylor Hall in 46 elements, and 41.07 MWD generated at General Atomics facilities in 56 elements (15 elements had burn from both facilities). With the exception of the fuel followers and standard fuel elements in positions Cl1 and D18, fuel in the B ring through the D ring (B01 through D17) had an average burn of 0.255 MWD per element, with a standard deviation of 0.037 MWD. With the exception of standard fuel elements in elements in core position F30, the standard fuel elements in the E and F rings and positions Cl and C18 had an average burnup of 0.710 MWD with a standard deviation of 0.026 MWD. The element in position F30 was an outlier, received from GA with almost a gram of 235U depleted. Fuel elements in the inner ring with the lower average burnup are marked on Fig. 4 with a
light shade of violet, and higher burn elements with a darker shade of purple. (Water voids are blue, graphite element's gray.)
.1
/1 Figure 4, Loading of Previously Burned Elements Control rod calibrations are accomplished under surveillance procedure SURV-6, and excess reactivity determinations are documented with SURV 3. The initial control rod reactivity worth calibration was completed on 03/31/1992, although excess reactivity was not determined until completion of initial testing in July when both SURV-3 and SURV-6 were performed. Surveillance data is provided in Table 1.
Table 1: 1992 UT TRIGA REACTIVITY ($), SURV-6 & SURV 3 PARAMETER 03/25/1992 07/23/1992 EXCESS NA 6.38 REG ROD 4.59 4.08 TR ROD 3.34 3.26 SHIM 2 3.46 3.30 SHIM 1 3.32 3.17 2.0 NUCLEAR PHYSICS MODELING Neutronic and gamma transport modeling for the UT TRIGA reactor is based on calculations with SCALE1 6.1. The SCALE code package is a comprehensive modeling and simulation suite for nuclear safety analysis and design developed under the sponsorship of the U.S. Nuclear Regulatory Commission (NRC) and maintained by Oak Ridge National Laboratory (ORNL). SCALE integrates a suite of programs and routines that pass, format, and process relevant information to support a variety of nuclear-system calculations based on standard nuclear data sets. SCALE control sequences select and integrate the programs and modules required for the type of calculation (and any selected options). For the UT TRIGA reactor, SCALE is used to calculate:
0 Critical mass Core Radial Peaking Factor Physics parameters & Flux density 1ORNL/TM-2005/39 Version 6.1, Scale: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design, Oak Ridge National Laboratories
Fuel Element Axial and Radial Peaking Factors Fission Product Poisons Transuranic buildup Fuel temperature reactivity deficit Water temperature reactivity deficit Control Rod worth Excess reactivity Shutdown Margin Effects of Experiments 0
Effects of Burnup Accident source Terms The SCALE Material Information Processor (MIP) is based on free-form engineering parameter input that specifies nuclear data to be selected from cross section libraries for use in control and functional modules. Physical modeling in SCALE is accomplished by the Generalized Geometry Package (SGGP);
SGGP uses a specific set of surface geometries as the framework, with volumes developed by defining orthogonal boundaries for surface extension. Analysis of the UT TRIGA reactor is based on the T-6 (TRITON) depletion sequence, calculating information of interest based on burnup and fission product generation. The T-6 sequence is based on a 3-D Monte-Carlo transport code, KENO-VI, in conjunction with an isotope buildup and decay code, ORIGEN. Other than (1) a broad description of the code and (2) identification of subroutines and codes used in processing, questions regarding methodology of calculation should be referred to the SCALE manual 2.
2.1 SCALE 6.1, T-6 Depletion Sequence The T-6 sequence is a specification in the TRITON computer code. TRITON is a multipurpose SCALE control module for transport, depletion, and sensitivity and uncertainty analysis with sequences to perform various calculations of interest. TRITON provides automated, problem-dependent cross-section processing followed by multigroup transport calculations. TRITON functions with the ORIGEN (Oak Ridge Isotope GENeration) depletion module to predict isotopic concentrations, source terms, and decay heat, as well as generating few-group homogenized cross sections for nodal core calculations. Isotopic concentrations are calculated by ORIGEN at burnup intervals specified in the input. The results of the ORIGEN calculations are used to modify materials specified in the input file for subsequent calculations.
A T-6 calculation involves three major steps. KENOVI, a 3-D Monte Carlo code, is used to determine region-specific multi-group fluxes, cross sections, and power generated by the material specified for depletion. From the information developed in the first step, COUPLE develops a "response function," a one-group cross section library. Using the one-group cross sections and the material power, ORIGEN performs depletion (fission and transmutation) calculations to determine material composition following the burnup interval. Material composition is transferred by KMART6 into KENOVI for the subsequent depletion calculation. The control modules and codes invoked by the T-6 depletion sequence include:
- 1.
CRAWDAD creating a continuous energy library for use by CENTRM PMC.
2 Scale: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design, ORNL/TM-2005/39, Version 6.1, June 2011. Available from Radiation Safety Information Computational Center at Oak Ridge National Laboratory as CCC-785
- 2.
BONAMI performing resonance self-shielding calculations for nuclides that have Bondarenko data associated with their cross sections.
- 3.
WORKER creating an AMPX working format library from a master format library.
- 4.
CENTRM using the pointwise continuous cross-section library and a cell description to create a pointwise continuous flux spectrum.
- 5.
PMC using the pointwise continuous flux spectrum created in CENTRM, to collapse pointwise continuous cross sections into a set of multigroup cross sections.
- 6.
CAJUN combining homogenized point cross-section libraries.
- 7.
CHOPS computing pointwise flux disadvantage factors and creates homogenized point cross sections.
- 8.
KENO-VI calculating keffof a 3-D system using the Monte Carlo method, and developing flux spectral flux distribution.
- 9.
KMART6 performing the flux post-processing for the KENO/VI sequence
- 10. COUPLE generating a one-group cross-section library for each depletion material from region-averaged multigroup cross sections and multigroup fluxes
- 11. ORIGEN calculating (in a matrix exponential expansion model) time-dependent concentrations, activities, and radiation source terms for a large number of isotopes simultaneously generated or depleted by neutron transmutation, fission, and radioactive decay.
- 12. OPUS program (optional) producing a condensed output file and data formatted for plotting from output generated by the ORIGEN-S code that computes reactor fuel depletion, activation and fission-product buildup, and radioactive decay.
2.2 SCALE OPTIONS IN TRIGA CALCULATIONS SCALE is a well-established tool for reactor analysis, with a set of default parameters to support commercial nuclear power reactors. Some default options create program conflicts in TRIGA analysis unless disengaged, and relaxing some of the defaults may improve processing time. Specification of parameters and the information desired for the UT TRIGA reactor leads to specific strategies.
2.2.1.
Default and Override Parameters Flux weighted, material averaged cross sections are calculated for the midpoint of burnup intervals. If there are large changes in materials over the burnup interval, the material averaged cross sections may not accurately represent the system cross section over the full range of the burnup interval. Because ORIGEN was originally designed to be used in power reactor systems operating at steady state power levels for very long periods of time, there is a default limit of 20 days on burnup intervals that ensure processed cross sections represent the material adequately. The burnup for a TIRGA reactor is generally not continuous or at a flux as high as a commercial reactor; therefore default may be overridden by substituting an alternate maximum number of days with the command "MAXDAYS=N."
Trace quantities of fission products may individually be low enough to not affect calculations, but in aggregate may have a significant effect. By default, TRITON automatically adds trace quantities of a set of nuclides to all fuel materials that have been determined to be important in the characterization of spent fuel. The nuclides are not explicitly listed in the material specification of the input file. The default set of added nuclides can be augmented to improve accuracy (with other defined groups of nuclides, up to about 380 total nuclides) using a command "ADDNUX=N" where N is a number that specifies sets of nuclides to be used. The default TRITON depletion sequences setting (N=2) adds 94 nuclides. Unfortunately, one of the nuclides in the augmented set is hydrogen. Since zirconium-hydrogen cross sections are explicit in the input file, hydrogen included automatically by ADDNUX
creates a conflict by using two different cross sections for a single nuclide. This conflict is alleviated by setting ADDNUX=0 and explicitly reproducing the augmented data set in the input, without hydrogen-i as a fission product. For fresh fuel, the atom-density of all additional isotopes set to a minimum (1E-20) results in a reactivity effect of about $0.04.
The number of input lines required to support fission product inventory is excessive. It is convenient to specify the isotopes in a separate file, with the file read into the SCALE input. A shell (with a copy command) at the beginning of the input file copies the file into the temporary directory. The material file is read into material specification with a right bracket and the file name. An example of the use of the shell and copy command to read the text file NEWFUEL.TXT into the temporary directory with the name FILE1.TXT is provided below; the read statement that is inserted in the material specification section is shown in a separate section.
=shell copy "%RTNDIR%\\NEWFUEL.TXT" "%TMPDIR%\\FUELI.TXT" end
<FUEL1.TXT The number of nuclides added to improve calculations increases processing time. In cases where parameters of interest are not related to burnup, the "INFDCUTOFF=X" specification compares the material cross section to the total system cross sections. If the system cross section is greater than the material cross section by the specified value, the material is not used in transport calculations.
Removing isotopes that have little impact relative to system cross section improves processing time.
ORIGEN was originally developed.for power level applications, and uses default mass normalization to a metric ton of heavy metal (i.e. uranium). While this is adequate for determining densities of materials such as fission products in large cores, it de-emphasizes fuel depletion in smaller cores. The "LENGTH="
option provides a means of normalizing to an arbitrary mass. This option was originally used with 2D models where scale is synonymous with length, and has been extended to 3D in the current version of SCALE. If the LENGTH is set to 1, normalization is the mass of the model based on fuel mass as calculated from volume and density.
2.2.2 Strategies Strategies developed for the UT TRIGA SCALE model are considered in three sections. First, strategies for acquiring data used in calculations are discussed. Second, strategies related to how the geometry is developed are discussed. Finally, development of the material specifications is discussed.
A.
Reporting and Calculations OPUS is a SCALE report module for specific parameters. The key word "symnuc" defines nuclides of interest. Parameters used in TRIGA calculations include material density (GPERCM is the specification for mass density, ATOMS the specification atomic density in units of atoms per barn-cm), mass (grams),
fractional material absorptions (ABSORB), radioactivity (CURIES), gamma spectrum (PHOTONS), and decay heat (WATTS). Multiple reports for different set of nuclides and parameters are generated using a "new case" delimiter within a single calculation. Application of the specifications is provided in Table 2A.
Table PURPOSE 235U, 238U mass burned Fuel material specification Reactivity Effects Loss of Pool Water Accident Source term Loss of Pool Water Accident Source term Fission Product Inventory MHA Source term 2A: OPUS Specifications Title s
URANIUM MASS u
MATERIAL SET 1 s
MATERIAL SET 2 s
MATERIAL SET 3 s
ymmnuc
-235 u-238 ee Table 2.B ee Table 2.C ee Table 2.D units grams atoms atoms atoms u-235 u-238 pu-238 pu-240 pu-241 pu-242 ABSORPTIONS xe-131 xe-133 xe-135 sm-147 sm-150 sm-151 sm-152 sm-153 DECAY HEAT u-236 pu-238 pu-watts 240 pu-241 pu-242 GSPEC CURIES time days days days days hours days For calculations used to generate material specification, nuclides specified in the ADDNUX routine (Table 4A/B/C) are reported; a limit on the number of nuclides prompted splitting the total set into three separate specifications. In modeling cores through operation, burnup is simulated to approximate the average power history for the core. The results of calculations of isotope concentrations following simulated operation are used as material specification for calculations of burned cores. Subsequent calculations are used to benchmark against reference reactivity values, characterize reactivity effects from burnup, or characterize the reactivity effect of specific materials (i.e., comparing keff from cases where the nuclide exists and is deleted).
Table 2B, MATERIAL SET 1 u-235 u-238 u-232 u-233 u-234 u-236 u-237 h-2 h-3 b-10 b-l1 n-14 n-15 o-16 o-17 f-19 p-31 s-32 i-127 i-129 i-130 i-131 i-135 w-182 w-183 w-184 w-186 y-89 y-90 y-91 zr-93 ag-107 ag-109 ag-111 al-27 am-243 as-75 au-197 ba-134 ba-135 ba-136 ba-137 ba-138 ba-140 be-9 bi-209 bk-249 br-79 br-81 cd-106 cd-108 cd-ll0 cd-lll cd-112 cd-il3 cd-1i4 cd-115m cd-116 ce-140 ce-141 ce-142 ce-143 ce-144 cf-249 cf-250 cf-251 cf-252 cm-241 cm-242 cm-243 cm-244 cm-245 cm-246 cm-247 cm-248 co-59 cs-133 cs-134 cs-135 cs-136 cs-137 dy-160 dy-161 dy-162 dy-163 Table 2C, MATERIAL SET 2 dy-164 er-166 er-167 es-253 eu-151 eu-152 eu-153 eu-154 eu-155 eu-156 eu-157 gd-152 gd-154 gd-155 gd-156 gd-157 gd-158 gd-160 ge-72 ge-73 ge-74 ge-76 he-3 he-4 hf-174 hf-176 hf-177 hf-178 hf-179 hf-180 ho-165 in-113 in-115 kr-78 kr-80 kr-82 kr-83 kr-84 kr-85 kr-86 la-139 la-140 li-6 Ii-7 lu-175 lu-176 mn-55 mo-100 mo-92 mo-94 mo-95 mo-96 mo-97 mo-98 mo-99 na-23 nb-93 nb-94 nb-95 nd-142 nd-143 nd-144 nd-145 nd-146 nd-147 nd-148 nd-150 np-237 pa-231 pa-233 pd-102 pd-104 pd-105 pd-106 pd-107 pd-108 pd-110 pm-147 pm-148 pm-148m pm-149 pm-iSl pr-141 pr-142 pr-143 pu-236 pu-238 pu-239 pu-240 pu-241
Table 2D, MATERIAL SET 3 pu-242 pu-243 pu-244 rb-85 rb-86 rb-87 re-185 re-187 rh-103 rh-105 ru-100 ru-101 ru-102 ru-103 ru-104 ru-105 ru-106 ru-96 ru-98 ru-99 sb-121 sb-123 sb-124 sb-125 sb-126 se-74 se-76 se-77 se-78 se-80 se-82 sm-144 sm-147 sm-148 sm-149 sm-150 sm-151 sm-152 sm-153 sm-154 sn-112 sn-114 sn-115 sn-116 sn-117 sn-118 sn-119 sn-120 sn-122 sn-123 sn-124 sn-125 sn-126 sr-84 sr-86 sr-87 sr-88 sr-89 sr-90 ta-181 ta-182 tb-159 tb-160 te-122 te-123 te-124 te-125 te-126 te-127m te-128 tc-99 te-120 te-129m te-130 te-132 th-230 th-232 xe-124 xe-126 xe-128 xe-129 xe-130 xe-131 xe-132 xe-133 xe-134 xe-135 xe-136 zr-95 Table 2E: Nuclides Identified in ADDNUX sets, Lacking Cross Section Data ti-50 ti-49 ti-48 ti-47 ti-46 th-234 th-233 th-229 th-228 th-227 sn-1 13 si-30 si-29 si-28 se-79 sc-45 ra-226 ra-225 ra-224 ra-223 pu-246 pb-208 pb-207 pb-206 pb-204 pa-232 np-239 np-238 np-236 np-2:35 ni-64 ni-62 ni-61 ni-60 ni-59 ni-58 mg-26 mg-25 mg-24 la-1 38 ir-193 ir-191 hg-204 hg-202 hg-201 hg-200 hg-199 hg-198 hg-196 ge-70 gd-153 ga-71 ga-69 fe-58 fe-57 fe-56 fe-54 es-255 es-254 er-170 er-168 er-164 er-162 dy-1 58 dy-1 56 cu-65 cu-63 cr-54 cr-53 cr-52 cr-50 co-58 cm-250 cm-249 cl-37 cl-35 cf-254 cf-253 ce-1 39 ce-1 38 ce-136 ca-48 ca-46 ca-44 ca-43 ca-42 ca-40 bk-250 be-7 ba-1 33 ba-132 ba-130 as-74 ar-40 ar-38 ar-36 am-244 am-242 ac-227 ac-226 ac-225 u-241 u-240 u-239 k-41 k-40 k-39 s-36 s-34 s-33 ag-11Om am-244m co-58m ho-166m The second and third group includes (1) hydrogen, (2) stable zirconium isotopes generated as fission products, and (3) a set of isotopes with inadequate library data (Table 2D). The hydrogen and zirconium isotopes are included in the calculations directly by standard materials specifications h-zrh2 and zr-zrh2, and the isotopes that do not have library data were excluded from material specifications.
The fraction of neutrons absorbed in fission product poisons is used to evaluate the impact of transuranic isotopes and fission products on performance. The activity and gamma spectrum of specific isotopes and the decay heat following shutdown was used to generate accident source terms.
KENO output includes calculations of fission density and the fraction of fissions in fissile materials of geometry units. Duplicating fuel elements with unique identification numbers for specific core locations is used provided data to calculate radial core-peaking factors. Dividing fuel element geometries axially is used to provide data to calculate core axial core peaking factors. Dividing fuel into concentric cylinders is used to provide data to calculate power distribution axially within a fuel element.
As previously discussed, modeling zirconium hydride based fuel requires atom-density specification, and it is convenient to specify fuel in a separate file. The file is called from the SCALE input file. The isotopes in the zirconium hydride are not altered, and these components can either be included in the input file or in the external file; in both cases the material is represented identically, but processing material files for subsequent operations can be somewhat simplified by including z-zrh2 and h-zrh2 directly in the SCALE input file.
While the T-6 sequence is designed to support material changes during operation, a single very short interval at very low power provides initial keff and keff following a small perturbation. This provides two sequential criticality calculations under essentially the same conditions, and provides an indicator for anomalies or issues in calculation.
Material temperature and/or density can be varied within a single calculation. This option provides data for reactivity calculations to determine temperature coefficients for fuel and water. An example of specification to perform calculations with varying water temperature is provided below. Temperature changes are specified by material (3 in the example). Density variations are specified by material identification (3 in the example), the number of isotopes in the material for which temperature is changed (2 in the example) and identification of the isotopes (1001 and 8016). Density is specified as the fraction of material theoretical density at the midpoint of the burnup interval.
READ TIMETABLE TEMPERATURE 3 0
293.15 1.OOE-09 293.15 2.OOE-09 297.15 3.OOE-09 301.15 4.OOE-09 305.15 5.OOE-09 309.15 6.OOE-09 313.15 7.OOE-09 317.15 8.OOE-09 321.15 9.00E-09 325.15 END DENSITY 3 21001 8016 0
1 1.00E-09 1 2.OOE-09 0.99909077 3.OOE-09 0.998025861 4.OOE-09 0.996816392 5.OOE-09 0.995471982 6.OOE-09 0.994000844 7.OOE-09 0.992410292 8.OOE-09 0.990706538 9.OOE-09 0.98889499 END END TIMETABLE B.
Geometry The geometry units used in UT TRIGA analysis are shown in Table 3. The table includes units used in all calculations ("Base Units"), and special-purpose units for specific calculations ("Optional Units"). A more complete description and use of geometry units is provided later.
Geometry specifications in SCALE allow "units" to be placed within an "array." With the core specified as an array bounded by the core barrel, locations within the core are filled with "units" containing fuel elements, "dummy" graphite rods, water, or control rods.
Table 3: Geometry Units Base Units Optional Units Unit #
Description Unit #
Description 1
Fuel element 8
Fuel Follower control Rod 2
Graphite element 200 Fuel axially/radially divided 3
Water void 201-206 Grid Plate Position BO1-B06 4
Fuel Follower control Rod 13 Fuel element, alternate burnup 5
Transient Rod 14 Fuel element, alternate burnup 6
Grid plate 7
Reflector and core
Variations in fuel specifications were used. Identifying individual elements (Units 201-206) allows determining local, element-specific flux and power. Segmenting fuel element geometry (Unit 200) allows determining flux profile in individual elements. As previously indicated, fuel in the 1992 core included 4 different burnup values; the single outlier is not modeled separately. Burnup values uses include new fuel in the fuel follower control rods (Unit 1), standard fuel elements located in the inner rings (Unit 13),
and standard fuel elements located principally in the outer rings (Unit 14).
The three fuel follower control rods are identical, and can be represented by a single unit description (Unit 4). Control rod positions are manipulated by translation along the Z axis. Duplicating the fuel follower control rod description with a different unit geometry number (Unit 8) allows a single control rod to be manipulated independently of other control rods in calculations.
C.
Material Specifications SCALE material specification uses standard (pre-defined) material identification or a combination of standard materials in a mixture, with mixture identification starting "wtp." Material identification is followed by an integer label, used in geometry units to specify the presence of the material. The next value in a material specification establishes density. For standard materials, a density-value greater than 0 is a multiplier applied to theoretical density. For "wtp" material, the density-value is the mass density. A density value of 0 indicates the next value in the specification is atom-density, with units of atoms per barn cm". For "wtp" materials, density is followed by a value indicating the number of standard materials in the mixture, then the standard material identification and density multiplier for each isotope. The last two values are the temperature and an "end" statement to indicate completion.
Standard compositions exist for all materials used in calculations for the UT TRIGA except for air and fuel. Air was specified as "wtpair." Fuel was specified by isotope atom density to support an implementation of fission products analogous to ADDNUX routines (with hydrogen removed).
Table 4: Sample Material Specification Standard Fraction No.
Iso.
Material Description StnadID Fato o
S.
ao-/m TM N
Material TD or p Iso.
Frac.
atom-B/cm TEMP END Zr in ZrH 2 zr-zrh2 1
0 0.034448473 300 END H in ZrH 2 h-zrh2 1
0 0.055117557 300 END Read file FUEL1.TXT
<FUEL1.TXT Zirconium Zirconium 2
1 300 END Water h2o 3
1 300 END Stainless Steel 304 ss304 4
1 300 END Graphite Graphite 5
1 300 END Aluminum aluminum 6
1 300 END Aluminum in RSR aluminum 7
0.944 300 END Wtptair 8
0.00123 2
300 END Air 7014 80 300 END 8016 20 300 END 1
300 END B4C Poison b4c 9
0.984 300 END Molybdenum Mo 10 1
300 END Materials used in UT TRIGA analysis are shown in Table 4, along with the required specification by column (where applicable). "Standard Material" indicates (except for the "read file" line) a pre-
defined label identified in the in SCALE standard material library. "ID" is the input file unique identification. "Fraction TD or p" is the fraction of thetorical density identified in the in SCALE standard material library or (for "wtp") density in g cm3. "No. Iso." and "Iso. Frac." is the number of isotopes used in a "wtp" specification and the fraction of each isotope in the material. If "Fraction TD or p" has a zero-value, the material concentration is atom density normalized to a barn. The TEMP and END specifications are self-explanatory.
3.0 UT CORE AND FUEL GEOMETRY Principle dimensions used in modeling the UT TRIGA core are taken from schematic drawings prepared by General Atomics. Major dimensions of fuel, core shroud, core barrel, and the reflector are incorporated in modeling. Although the models include beam ports and the rotary specimen rack geometry, modeling of these features is not discussed in this section. Fuel end fittings are modeled as conical, stainless steel structures at the ends of the element (not including the fins or end-pins). End fittings for graphite "dummy" elements are not modeled.
3.1 CORE BARREL (REFLECTOR AND GRID PLATES)
The radial reflector is annular, with a cylindrical outer surface (radius of 81.994 cm) and the inner surface conforming to the inner shroud. For convenience the reflector is assumed to extend the full height of the inner shroud; however, the reflector actually ends about 3 in. above the lower grid plate (T2W210J111), at an elevation corresponding to about Y2 of the fuel element's lower axial reflector (about 1 % in. below the fuel). The reflector annulus (core barrel) and the lower grid plate are composed of aluminum alloy 6061. They are shaped by two hexagonal prisms, one rotated 30° as illustrated in Fig.
- 5. One of the hexagonal shapes is 55.625 cm, the other 52.637 cm (GA Technologies Inc. drawing T2W210J111). The lower grid plate is 1.25 in, thick (NETL BGPO01), with the upper surface 33.249 cm, and the lower surface 36.424 cm below the center of the fuel.
Figure 5, Core Shroud and Lower Gird plate The annulus at the upper part of the inner reflector surface and the upper grid plate are cylindrical. The upper reflector has a space for a rotary specimen rack, displacing the reflector to an outer radius of
35.715 cm (Rotary Rack Assy. Mark I & II, TO6S14E115). Center to center distance for the fuel element positions (pitch) in the grid plates is 1.714 in. (4.355 cm, T2W210E108 - Top Grid Plate, NETL BGPOO01 -
Bottom Grid Plate). Upper grid plate penetrations are nominally 1.505 in. (3.823 cm) in diameter. The bottom grid plate has a set of positions with the same diameter as the top, shaded in Fig. 6, and the remaining positions 1.250 in. in diameter. In addition to the variations in lower grid plate penetration diameters, the upper grid plate has clearance for fuel elements, graphite elements, control rods, control rod guide tubes, and experiments. To simplify modeling, all grid plate penetrations are assumed at the same diameter as the component inserted in the position.
The upper grid plate is 5/8 in. thick (Top Grid plate, NETL TGPOO01), with the lower elevation and the upper elevation referenced to the center of the fuel 42.799 cm and 41.224 respectively. Dimensions of penetrations in the upper grid plate are provided in Table 5.
Table 5: Upper Grid Plate Dimensions Label A
Description In.
B3 B
Fuel Position 1.505 L-C 6/7 Element Facility diameter 5.140 1-8 D
3 Element Facility Y displacement 4.285 E
3 Element Facility X displacement 3.464__
F 3 Element Facility displacement 5.999 I* -M G
3 Element Facility X displacement 3.464 H
Grid Plate Diameter 21.75
'JK 1
6/7 Element, Diameter 1 4.002 J
Grid plate thickness 0.62 K
6/7 Element Facility, Diameter 2 4.175
__-H-L 3 Element Cutout 2.370 M
Alignment pin hole 5/6 Figure 6, Upper Grid plate Although the core is not actually symmetric about the fuel, the reference point selected for physics models is the center of the active fuel. The distance from the center of the fuel to the top of the lower grid plate is 13.09 in. (33.249 cm) above the top of the lower grid plate (based on dimensions from GA drawings3). Modeling of standard fuel elements as based on these dimensions is provided in Table 5.
3.2 TRIGA FUEL ASSEMBLIES TRIGA fuel is composed of stainless steel cladding enclosing three 5 in. cylinders of uranium in zirconium hydride. Standard TRIGA fuel is identified4 as 8.5 to 12 wt % uranium (20% enriched) as a fine metallic dispersion in a zirconium hydride matrix. The H/Zr ratio is nominally 1.6 (in the face centered cubic delta phase).
3 TS13S210B217, TS13S210B229, TS13S210C212, TS13S210C214, TS13S210C218, TS13S210C226, TS13S210C227, TS13S21OD210, and TS13S210D213, derived by GA 4 GA Project No. 4314, The U-ZrH, Alloy: Its Properties and Use in TRIGA Fuel, M. T. Simnad (Feb. 1980) and NUREG 1282, Safety Evaluation Report on High-Uranium Content, Low-Enriched Uranium-Zirconium Hydride Fuels for TRIGA Reactors, Docket No. 50-163 (Aug 1987)
Fuel used at the UT TRIGA is nominally 8.5 wt%, 38 grams 235U. Density is calculated from fuel mass and dimensions. At nominal enrichment of 20% with fuel volume 389.5 cm 3, fuel has 2235.3 g U-ZrH1.6, with density calculated at 5.74 g/cm 3; at enrichment of 19.5%, fuel mass is calculated to be 2292.6 g U-ZrH1.6, with density 5.89 g/cm 3. Axial graphite reflectors are installed above and below the fuel, with a protective molybdenum disk between the lower reflector and the fuel. A gap above the upper axial reflector permits thermal expansion and provides space for outgassing of fission products and hydrogen.
End fittings are welded to the cladding above the gas gap and below the lower axial reflector; the end fittings are machined to extend into the cladding.
Table 6: TRIGA Standard Fuel Element (SFE) Dimensions
-35
-6
-8 NODE 7
1 Top of upper pin 2
Top of upper fins 3
Top of reinforcment Bottom of reinforcment Top of conical shape Bottom of conical shape 5
Top of cladding Weldment plug top Weldment plug bottom Top of gap Bottom of gap Top of axial reflector Bottom of axial reflector Top of fuel Bottom of fuel Top of Moly disk Bottom of moly disk 10Top of axial reflector Bottom of axial reflector Top of weldment plug Bottom of welment plug 12 Bottom of cladding Top of conical end-shape Top of lower fins 13 Top of lower grid plate Start fin taper Bottom of conical shape Top of lower pin 15 Bottom of lower pin Z2 in.
15.50 13.94 13.06 12.81 10.81 10.56 10.060 7.50 cm 39.370 35.408 33.172 32.537 27.457 26.822 25.552 19.050
-7.5
-19.050
-9 13 15 Figure 7, SFE
-7.531
-11.251
-11.501
-13.09
-19.129
-28.578
-29.213
-33.249
-13.695
-34.785
-14.441
-36.680 Although the end fittings are integral subassemblies, they are modeled for convenience as a plug located within the cladding and a separate conical shape above (below) the upper (lower) part of the cladding. The fins and the upper and lower pins (Fig. 7, 8) extending beyond the conical sections are neglected in modeling; the low mass of the fins and the distance (of the fins and pins) from active fuel should result in negligible impact on core physics performance.
cuff"w(SUw~rg 0"0"e $1dwjijPtw o
Ad@Mr Figure 8, Standard Fuel Element Details Early TRIGA reactor fuel was supported in the core on a pin at the lower end of the fuel element that rested in a depression in the lower grid plate; later TRIGA fuel is supported by three fins in contact with the lower grid plate. An adapter (Fig. 9) is required to use older fuel in current TRIGA grid plates. The adapter mimics the geometry of the fins structure, and the older fuel with the adapter is not differentiated from the newer fuel in modeling.
There are 9 grid plate locations designed to accommodate control rods and the central thimble, with holes at the same diameter in the upper and lower grid plates. Fuel can be used in these locations using an adapter (a hollow cylinder) secured to the aluminum "safety plate" below the lower grid plate. The top of the adapter mimics the top of the lower gird plate, providing support surface for fuel elements fins.
Holes in the surface of the adapter permit cooing flow through the bottom of the fuel element (Fig. 9). These adapters are designed to make the geometry of the lower grid plate in these Figure 9 positions compatible with the geometry of the other penetrations, and are not modeled. A Adapter photograph of the area under the lower gird plate is provided in Fig. 10.
t-igure Iu, Area beiow tne Lower (rla Plate
Table 6: Standard Fuel Element Modeling Component Upper Lower Az in.
Az cm r, cm r2 cm o
NODE NODE A
Upper conical shape 3
5 2.25 5.715 0
1.8771 B
Upper weldment Plug 5
6 0.25 0.635 1.8263 na C
Air gap 6
7 0.5 1.27 1.8263 na D
Upper aixial relfector 7
8 2.56 6.502 1.8263 na E
Fuel 8
9 15 38.1 1.8263 na F
Cladding 5
12 22.32 56.570 1.8263 1.8771 G
Moly disk 9
10 0.031 0.079 1.8263 na H
Lower axial reflector 10 11 3.72 9.449 1.8263 na I
Lower weldment plug 11 12 0.25 0.635 1.8263 na Figure 11, J
Lower conical shape 12 14 2.194 5.572 0
1.8771 SFE Model Dimensions of UT TRIGA reactor fuel assemblies for computational models are provided in Table 5 and Fig. 7, based on dimensions from GA drawings identified in private communication with General Atomics. A simplified model for calculations is provided in Table 6 and Fig. 11.
3.3 CONTROL RODS Two types of control rods are used in the UT TRIGA: fuel follower control rods (FFCR) and a transient control rod. The poison sections of control rods are composed of B4C, heat pressed to a densitys greater than 2.48 g cm 3 (theoretical density of B4C is 2.52 g cm3). Fuel follower control rods incorporate a section of TRIGA fuel below the control element so that as boron poison is removed from the core, fuel is added to the core. The transient control rod uses a guide tube and has an air filled follower. Control rod components are positioned by z-axis translation to simulate control rod motion.
3.3.1 FUEL FOLLOWER CONTROL RODS.
Standard fuel element follower control rods are stainless steel tubes with welded end fittings approximately 45 in. (114 cm) long by 1.35 in. (3.429 cm) in diameter (1991 UT SAR). Component dimensions are taken from GA drawing TOS250D225 (Control Rod - Fuel Follower).
Table 7: FFCR Component Thickness (Z Axis)
Component in.
cm Upper End Fitting 1.5 3.81 Upper Air Void 3.5 8.89 Magneform Plug 0.5 1.27 Poison/Air Gap 0.12 0.3048 Poison 15 38.1 Magneform Plug 0.5 1.27 Fuel/Air Gap 0.25 0.635 Fuel 15 38.1 5 GA Drawing TOS250B226, Poison
Magneform Plug Lower Air Void Lower End Fitting 1
5.375 0.5 2.54 13.6525 1.27 The upper 6.5 in. (16.51 cm) of the fuel follower control rod is an air void, separated and secured from the poison section by a 0.5 in. (1.27 cm) plug secured with a magneform weld. There is a small (0.12 in, 0.305 cm) air gap at the top of the poison section, between the poison and the plug. The poison is 15 in.
(38.1 cm) of B4C. 1.187 in. (3.01498 cm) in diameter. The poision section is separated from the fuel follower section by another 0.5 in. plug and magneform weld. The top of the fuel follower section is a 0.25 in (0.635 cm) air gap, above the fueled. The fuel rests on a double thickness 1 in. (2.54 cm) plug and magneform weld, followed by a 6.5 (16.51) in. air void. The bottom air void has an aluminum insert with wall thickness 0.35 in. (0.089 cm). For convenience, the reference point for the control rod in Table 8 aligns the fuel section with the center of the fuel in standard fuel elements when the rod is withdrawn.
Table 8: Fuel Follower Control Rod Geometry KEY DESCRIPTION ELEVATION in cm A
Top upper end 13.87 35.230 B
Bottom upper end 12.37 31.420 Top of top void C
Bottom of top void 8.87 22.530 Top of top plug D
Bottom of mag plug 8.37 21.260 Top of poison gap E
Bottom of poison gap 8.25 20.955 Top of poison Bottom of poison F
Top of lower poison mag
-6.75
-17.145 plug G
Bottom of lower mag plug
-7.25
-18.415 top of fuel gap H
bottom of fuel gap
-7.5
-19.05 top of fuel H
bottom of fuel top of lower mag plug bottom of lower mag plug
-23.5
-59.69 top of lower air follower K
bottom of air follower top of lower end fitting L
hnttnm nf Inwagr anti fittina
-70 1r
-"7A r1 2 Figure 11, FFCR 5
3.3.2 TRANSIENT CONTROL ROD.
The transient control rod is less complex than the fuel follower control rod. Transient rod cladding is a 1.25 in. (3.175 cm) diameter aluminum tube with wall thickness 0.028 in. (0.071 cm) 6. The transient rod operates within an aluminum guide tube (1.490 in. or 3.7846 cm in diameter machined from 1.5 in./3.81 cm aluminum tubing with a wall thickness 0.065 in. or 0.1651 cm 7; therefore inner diameter is 3.480 cm) secured to the upper grid plate and the safety plate (below the lower grid plate). The guide tube is perforated by Y2 in. holes at 900 rotations on 1 in. centers at the top and bottom of the core barrel. The guide tube extends 5 feet above the safety plate, below the lower grid plate. Perforations and extensions above the upper-and below the lower-grid plate are not modeled.
The poison section of the transient rod is 15 in (38.1 cm)s. A double (1 in., 2.54 cm) plug with a magneform weld secures the poison section at the top and bottom. The air follower above the poison section is in an assembly 5.94 in., which includes the upper end fitting and the upper magneform weld.
The air follower under the plug is 20.88 in. (53.02 cm) long, terminating in a bottom end-fitting.
Table 9: Transient Rod Dimensions KEY DESCRIPTION ELEVATION8 in cm End A
Fitting B
134C Magneform Weld "D
Air Follower End iE Fitting "
j F Figure 12, Transient Rod A
Top of end fitting Bottom of end fitting B
Top of poison section Bottom of poison section Top of magneform weld Bottom of magneform D
weld Top of air follower Bottom of air follower Top of end fitting F
Bottom of end fitting 9
22.860 7.5 19.050
-7.5
-8.5
-19.050
-21.59
-27.75
-70.485
-28.25
-71.755 Figure 13, Guide Tube 4.0 MODELING As previously described, SCALE uses a standard description of geometry based on extruding planer surfaces from a base to an upper extension. A material is associated with each of these volumes. Details of the geometry used to model the UT TRIGA are followed by a description of the materials.
4.1 GEOMETRY UNITS Component geometries are used to develop "units" within the SCALE SGGP module. The model is bounded by a "global unit" composed of the reflector and core. The reflector portion includes the radial graphite reflector, the reflector canister, beam port tubes, rotary specimen rack, grid plates, and water.
Each unit has one assembly identified as the boundary, a volume that encompasses all geometries in the unit.
6 T0S252D191-Transient Rod Assembly (ADJ) 7 T135210D150, Guide Tube - Transient Rod 8 T0S252D191 & 1991 UT SAR
The core portion is an array statement defining a set of hexagonal prisms bounded by the inner shroud.
The hexagonal prisms are filled with appropriate geometry units that incorporate components inserted in core spaces. Components that fill the core array include cells with grid plate penetrations, water, and (where appropriate) fuel elements, graphite elements, water voids, fuel follower control rods, or the transient rod. One unit is composed of two solid aluminum plates at the thickness of the grid plates separated by water. Fig 14. illustrates three hexagonal positions filled with fuel. Fig. 15 is an example of a core configured with 71 standard fuel elements and 3 fuel followers with non-fuel positions occupied by graphite rods, and with non-fuel positions occupied by water filled channels (or water voids).
0r*,
0J 30S2 4, Figure 14, Dimensions of Contiguous Unit Cells with Fuel 6
6 6
6 6
6 6
6 6
6 6
6 6
6 6
6 6
6 6
6 6
6 6
6 6
6 6
6 6
6 6
6 6
6 6
2 2
2 2
2 6
6 6
6 6
6 6
6 6
6 3
3 3
3 3
6 6
6 6
6 6
6 6
6 2
6 6
6 6
6 6
6 6
6 6
6 6
23 6
6 6
2 6
6 6
6 3
36 6
6 6
6 23 3 6 6
6 6
2 6
6 6
6 3
3 6
6 6
6 2
2 6
6 6
3 13 6
6 6
3 6
6 6
r 3
66 66 6
2 2
6 6
6 6
3 3
6 6
6 6
2 2
6 6
6 6
6 6
3 A3 6
66 66 6
6 2
4 2
6 6
6 6
63 6
3 6
6 6
6 6
6 6
6 2
2 2
2 26 6
6 6
6 6
6 3
6 6
3 3
3 3
3 6
6 6
6 6
6 6
6 6
6 6
6 6
6 6
6 6
6 6
6 6
6 6
6 6
6 6
6 6
6 6
6 6
6 6
Figure 15, Core Array with Fuel (1) Graphite (2), Water void (3), Control Rods (4 & 5) & Grid Plates (6) 4.1.1 Fuel Geometry Unit The SCALE model description for a unit to be inserted in a grid plate position that contains TRIGA fuel is provided in Table 10. As noted in Table 3, the fuel unit may be labeled or segmented to support specific calculations. When it is desired to determine power in a specific element (or set of elements, such as the B ring), duplicate units with unique labels are used in the core array. If materials are different
between elements, e.g. generated by different power histories, the unit can be duplicated with a different fuel material. When fission distribution within an element is significant, the fuel cylinder is segmented either radially or axially, as appropriate. The boundary for unit I is hexprism 30.
Table 10: unit 1: com="unit 1: fuel channel" (all units cm)
Description Geometry ID Radius or Pitch Top Radius Bottom Full Channel hexprism 30 2.177 32.309
-36.424 Grid Plate hexprism 40 2.177 32.309 30.734 Grid Plate H20 cylinder 41 1.8771 32.309 30.734 End Fitting /2 cone 10 0.635 32.537 1.8771 27.457 Cladding cylinder 11 1.8771 27.457
-29.213 End Fitting Y2 cylinder 12 1.8263 27.457 26.822 Gas Gap cylinder 13 1.8263 26.822 25.552 Axial Reflector cylinder 14 1.8263 25.552 19.05 Fuel cylinder 15 1.8263 19.05
-19.05 Zirc Filler cylinder 16 0.285 19.05
-19.05 Moly Disk cylinder 17 1.8263
-19.05
-19.129 Axial reflector cylinder 18 1.8263
-19.129
-28.578 End Fitting Y2 cylinder 19 1.8263
-28.578
-29.213 End Fitting Y2 cone 20 1.8771
-29.213 0.635
-34.775 Grid Plate hexprism 50 2.177
-33.249
-36.424 Grid Plate H20 cylinder 51 1.8771
-33.249
-36.424 4.1.2 Graphite Dummy Rod Geometry Unit The SCALE model description for a unit to be inserted in a grid plate position that contains a dummy (graphite) rod is provided in Table 11. The boundary for unit 2 is hexprism 20.
Table 11: unit 2 com="unit 2: graphite rod "(all units cm)
Description Geometry+
ID Radius top bottom or Pitch Channel Hex-prism 20 2.177 32.309
-36.424 Grid Plate H20 Cylinder 25 1.873 32.309
-36.424 Grid plate Hex-prism 30 2.177 32.309 30.734 Upper Pin Cylinder 40 0.635 32.537 27.457 Cladding Cylinder 50 1.873 27.457
-29.213 Upper Plug Cylinder 51 1.822 27.457 26.822 Graphite Cylinder 52 1.822 26.822
-28.578 Lower Plug Cylinder 53 1.822
-28.578
-29.213 Lower Pin Cylinder
.60 0.635
-29.213
-34.775 Grid Plate Hex-prism 70 2.177
-33.249
-36.424 4.1.3 Water Void Geometry Unit The SCALE model description for a unit to be inserted in a grid plate position that does not contain a rod (dummy or fuel) is provided in Table 12; this is considered a water-void. The boundary for unit 3 is hexprism 30.
Table 12: unit 3 com="unit 3: Water Channel" (all units cm)
Radius Description Geometry ID Radius top bottom or Pitch Channel Hex-prism 30 2.177 32.309
-36.424 Grid plate Hex-prism 40 2.177 32.309 30.734 Grid plate H20 Cylinder 41 1.911 32.309 30.734 Grid Plate Hex-prism 10 2.177 32.309
-36.424 Grid Plate Hex-prism 50 2.177
-33.249
-36.424 Grid plate H20 Cylinder 51 1.922
-33.249
-36.424 4.1.4 Fuel Follower Control Rod Geometry Unit The SCALE model description for a unit to be inserted in a grid plate position that contains a standard, fuel follower control rod is provided in Table 13. Components in the table shaded and bounded by the dashed line are modeled with an axis translation within the geometry unit to simulate withdrawal and insertion of the control rod by specifying a position on the vertical axis. Where motion is required for a single control rod, the geometry unit is duplicated with a different unit number with the alternate unit number placed to represent the single rod. The boundary for unit 1 is hexprism 60.
Table 13: Unit 4, Standard Control Rod 9 (all units cm)
Description Geometry ID Radius Top Bottom Channel hexprism 60 2.177 32.309
-36.424 Grid plate hexprism 50 2.177 32.309 30.734 Cladding cylinder 30 1.715 35.23
-74.613 UpperVoid cylinder 10 1.6637 35.23 22.53 Magneform Plug cylinder 11 1.6637 22.53 21.26 Gap cylinder 12 1.6637 21.26
-17.145 B4C cylinder 14 1.651 20.955
-17.145 Magneform Plug cylinder 15 1.6637
-17.145
-18.415 Gap cylinder 16 1.6637
-18.415
-19.05 Fuel cylinder 17 1.6637
-19.05
-57.15 Zirc Filler Magneform Plug Lower void End fitting Grid Plate cylinder 18 0.285
-19.05
-57.15 cylinder 19 1.6637
-57.15
-59.69 cylinder 20 1.6637
-59.69
-73.343 cylinder 25 1.6637
-73.343
-74.613 hexprism 40 2.177
-33.249
-36.424 4.1.5 Geometry Unit for Transient Control Rod The SCALE model description for a unit to be inserted in a grid plate position that contains the transient control rod is provided in Table 14. Movement is simulated similar to the fuel follower control rod. The boundary for unit 5 is hexprism 30.
9 Items in the boxed are specified as "origin x=0 y=0 z=0" for rod fully out, origin x=0 y=0 z=38.1" for rod fully out
Table 14: Unit 5, Transient Control Rod 8 (all units cm)
Description Geometry ID Radius Top Bottom Channel hexprism 30 2.177 32.309
-36.424 Outer Guide Tube cylinder 60 1.911 32.309
-36.424 Inner Guide Tube cylinder 65 1.829 32.309
-36.424 Grid plate hexprism 40 2.177 32.309 30.734 Cladding cylinder 20 1.59 32.309
-76.248 End Fitting cylinder 10 1.519 32.309 29.151 Air void cylinder 11 1.519 29.151 20.828 Magneform Weld cylinder 1.2 1.519 20.828 19.05 Radial Gap cylinder 13 1.519 19.05
-19.05 B4C cylinder 14 1.3 19.05
-19.05 Magenform Weld cylinder 15 1.519
-19.05
-20.32 Air void cylinder 16 1.519
-20.32
-64.61
_ _End Fitting................. cylinder ------- 17.... 1.519
-64.61 --------
66.:
Grid plate hexprism 50 2.177
-33.249
-36.424 4.1.6 Geometry Unit for Solid Grid Plate The core grid plate does not have penetrations at the periphery of the core. Since the core is represented by an array of hexagonal spaces, the spaces at the core periphery are solid aluminum plates separated by water. The SCALE model description for these grid plate positions is provided in Table 15.
The boundary for unit 1 is hexprism 30.
Table 15: Unit 6, Grid Plate and Water (all units cm)
Description Geometry ID Radius Top Bottom Grid Plate hexprism 30 2.177 32.309
-36.424 Water hexprism 40 2.177 32.309 30.734 Grid Plate hexprism 50 2.177
-33.249
-36.424 4.1.7 Geometry Unit for Reflector and Core The reflector and core includes an aluminum "shroud" composed of two nested hexagonal geometries, the rotary specimen rack, the graphite reflector, and beam ports. The beam ports require translation not shown. The SCALE model description for the reflector and core is provided in Table 16. The boundary for unit 7 is cylinder 40.
Table 16: Global Unit Unit 7, Reflector and Core1° (all units cm)
Description Geometry ID Radius Top Bottom Core Barrel Hex rhexprism 10 25.161 2.309
-36.424 Core Barrel Hex rhexprism 11 25.796 2.309
-36.424 Rotary Specimen Rack cylinder 20 30.083 32.309 6.509 Rotary Specimen Rack cylinder 21 30.718 32.309 6.509 10 Core barrel hexagonal prisms are rotated by "rotate al=0 a2=0 a3=60"; beam ports are translated by (50) origin x=0 y=35.255 z=-6.985, (51) origin x=-33 y=1O z=-6.985 rotate al=O a2=0 a3=-30, (52) origin x=0 y=0 z=-6.985, and (53) origin x=O y=O z=-6.985 rotate al=O a2=O a3=60
Table 16: Global Unit Unit 7, Reflector and Core10 (a/l units cm)
Description Geometry ID Radius Top Bottom Rotary Specimen Rack cylinder 22 31.115 32.309 6.509 Rotary Specimen Rack cylinder 23 35.401 32.309 6.509 Rotary Specimen Rack cylinder 24 36.671 32.309 6.509 Outer Reflector Surface cylinder 40 59.69 32.309
-36.424 Beam Port xcylinder 50 7.62 90
-90 Beam Port ycylinder 51 7.62
-10
-90 Beam Port ycylinder 52 7.62 0
-90 Beam Port ycylinder 53 7.62 0
-90 4.2 MATERIALS ORNL/TM-2005/39, Section M.8, Standard Composition Library, provides material composition data for all materials used in modeling with SCALE except for air and the zirconium hydride fuel. Nominal values of density from the Standard Composition Library are used except in the case of the control rod poison sections and aluminum in the rotary specimen rack that have complex structures (approximated as 94.4% theoretical density). As previously note, General Atomics specifications on boron carbide used in the control rods is at least 98% of theoretical density listed in ORNL/TM-2005/39, Table M8.2.4. Small volumes of air are modeled in gas-gaps for the fuel, void spaces in control rods, and beam port tubes, assumed to be 80% nitrogen and 20% oxygen at 0.00123 g/cm 3 with only the major stable isotopes.
4.2.1 Material Specification in Geometry Unit 1 (Fuel)
Fission product and transuranic isotopes in a fuel element are functions of initial loading and power history. Benchmarking to actual conditions requires some level of rigor in defining materials representative of initial conditions. In this case, calculations are performed to determine isotope concentrations at burnup intervals. Geometry unit 1 (Table 17) is designated to use fresh fuel, and is reproduced under different unit labels (13 or 14) where a material represents fuel composition at burnup intervals as calculated by simulated operation at a specified power for a specified time.
Table 17: Material Specification in Geometry Unit 1 (Fueled Channel)
Description Material ID Geometry Included Excluded Grid Plate Aluminum 7
1 40 51 Grid Plate Water Water 3
1 41 10 End Fitting 1/2 SS304 4
1 10 Cladding 5S304 4
1 11 12-1' End Fitting 1/2 SS304 4
1 12 Gas Gap Air 9
1 13 axial reflector Graphite 5
1 14 FUEL Fuel 111 1
15 16 Zirc Filler Zr 2
1 16 9
11 Alternately 13 or 14, as required
Table 17: Material Specification in Geometry Unit 1 (Fueled Channel)
Description Material ID Geometry Included Excluded Moly disk Molybdenum 12 1
17 axial reflector Graphite 5
1 18 End Fitting 1/2 SS304 4
1 19 End Fitting 1/2 SS304 4
1 20 Grid Plate Aluminum 7
1 50 51,20 Grid Plate Water Water 3
1 51 20 Remainder Water 3
1 30 Everything else 4.2.2 Material Specification in Geometry Unit 2 (Graphite Rod Channel)
Graphite "dummy" rods are used in core positions, with material specification provided in Table 18.
Table 18: Material Specification in Geometry Unit 2 (Graphite Rod)
Description Material ID Geometry Included Excluded Channel water Water 3
1 20 25,30,70 Upper Grid Plate Aluminum 7
1 30 25 Grid plate water Water 3
1 25 40,50,60 Upper Pin Aluminum 7
1 40 Cladding Aluminum 7
1 50 51-53 Upper plug Aluminum 7
1 51 Graphite Graphite 6
1 52 Lower Plug Aluminum 7
1 53 Pin Aluminum 7
1 60 Lower Grid Plate Aluminum 7
1 70 25 4.2.3 Material Specification in Geometry Unit 3 (Water Void Channel)
Grid plate positions may be left vacant. Therefore this unit contains upper and lower gird plates and water. The unit is referred to as a "water-void," described in Table 19.
Table 19: Material Specification in Geometry Unit 3 (Water Void Channel)
Description Material ID Geometry Included Excluded Upper Grid Plate Aluminum 7
1 40 41 Upper Grid Plate Hole Water 3
1 41 Water Water 3
1 30 40,41,50,51 Lower Grid Plate Aluminum 7
1 50 51 Lower Grid Plate Hole Water 3
1 51
4.2.4 Material Specification in Geometry Unit 4 (Fuel Follower Control Rod)
The material specification for fuel follower control rods is provided in Table 20. The fuel follower control rods in the 1992 core were new, requiring material specification as fresh fuel.
Table 20: Material Specification in Geometry Unit 4 (Fuel Follower Control Rod)
Description Material ID Geometry Included Excluded Channel Water Water 3
1 60 50,40,30 Upper Grid Plate Aluminum 7
1 50 30 Cladding SS304 4
1 30 10,11,12,15,16,17,19,20 Upper Void Air 9
1 10 Magneform Weld SS304 4
1 11 Gas Gap Air 9
1 12 14 Poison B4C 10 1
14 Magneform Weld SS304 4
1 15 Gas Gap Air 9
1 16 FUEL Fuel 1
1 17 18 Zirc Filler Zirconium 2
1 18 Magneform Weld SS304 4
1 19 Lower Void Air 9
1 20 Lower Grid Plate Aluminum 7
1 40 30 4.2.5 Material Specification in Geometry Unit 5 (Transient Rod)
The material specification for the transient control rod is provided in Table 21.
Table 21: Material Specification in Geometry Unit 5 (Transient Rod)
Description Material ID Geometry Included Excluded Channel Water Water 3
1 30 40,50,60 Guide Tube Aluminum 7
1 60 65 Water in Guide Tube Water 3
1 65 20 Grid Plate Water Aluminum 7
1 40 60 Cladding Aluminum 7
1 20 10,11,12,13,15,16,17 End Fitting Aluminum 7
1 10 Upper Void Air 7
1 11 Magneform Weld Aluminum 7
1 12 Air Surrounding Poison Air 9
1 13 14 Poison B4C 11 1
1 15 Lower Void Air 9
1 16 Lower End Fitting Aluminum 7
1 17 Lower Grid Plate Aluminum 7
1 50 60
4.2.6 Material Specification in Geometry Unit 6 (Grid Plate and Water)
The material specification for array locations where the grid plate does not have penetrations (i.e., areas at the periphery of the core barrel) rods is provided in Table 22.
Table 22: Material Specification in Geometry Unit 6 (Grid Plate and Water)
Description Material ID Geometry Included Excluded Upper Grid Plate Aluminum 7
1 40 Lower Grid Plate Water 3
1 30 40,50 Lower Grid Plate Hole Aluminum 7
1 50 4.2.7 Material Specification in Geometry Unit 7 (Core and Reflector)
Outer surfaces of the reflector bound the model. The material specification for the core and reflector (including the rotary specimen rack and the beam ports) is provided in Table 23.
Table 23: Material Specification in Geometry Unit 7 (Core and Reflector)
Description Material ID Geometry Included Excluded Core Barrel Aluminum 7
1 11 10 Reflector Graphite 6
1 40 24,11,50,51,52,53 Beam Port Air 9
1 50 40,11 Beam Port Air 9
1 51 40,11 Beam Port Air 9
1 52 40,11 Beam Port Air 9
1 53 40,11 RSR Air 9
1 20 11 RSR Walls Aluminum 7
1 21 20 RSR Walls Aluminum 7
1 22 21 RSR Tubes Aluminum 2 8
1 23 22 RSR Walls Aluminum 7
1 24 23 5.0 LIMITING CORE CONFIGURATION NUREG 1537 requires a Limiting Core Configuration (LCC) be identified as the configuration that has a fuel element with the highest power density. The highest power density occurs in the fuel element where the ratio of the power in an element to the average power across all elements (power per element) is largest.
NUREG 1537 requires information relative to the operational core. Fuel loading in excess of the limiting core configuration supports compensates for fuel burnup and the negative reactivity that may occur with experiment insertion. Operational core loading is limited by the maximum excess reactivity permitted in the Technical Specifications.
5.1 Determining the Limiting Core Configuration The highest average power per fuel element will occur when the number of fuel elements is smallest.
The use of a reflector reduces critical mass by reducing neutron leakage. With less leakage, the power generated in fuel elements at the core periphery will be higher and generally reduce the peak-to-average power ratio. However, fewer fuel elements are required if non-fuel spaces are filled with graphite rods; the use of graphite rods will result in higher average power (per element). Calculations were therefore performed to determine the minimum number of fuel elements required to achieve criticality using graphite reflector rods and also using water voids.
The highest power density will occur when the reactor is operating at full power. Therefore, SCALE T-6 calculations were performed with the number of fuel elements varying from 49 to 117 assuming fresh fuel at varying fuel temperatures (300 K, 350 K, 450 K, 550 K and 600 K) simulating fuel temperature range from zero-power critical to full power operation.
The highest power density will occur in an element closest to the center of the core. Therefore the B-ring elements were individually labeled to provide measurement of fissions in each element.
The average power for each fuel element is determined based on the minimum number of fuel rods required to support full power operation for graphite and for water void core configurations. The ratio of power generated in each B ring element to the average power per element is determined for the two cases. The power in each B ring element is determined as the product of the average power and the peaking factor for that B ring element.
Therefore, calculations using the base units indicated in Table 3 and alternate units 201-206 were performed to determine the fuel element producing the maximum power assuming:
The number of fuel elements was varied from 49 elements to 117 elements (Table 24)
Two sets of core configurations with gird plate positions that are not occupied by fuel or control rods specified as:
allwatervoids all graphite rods Fuel temperatures varying from room temperature to 600 K B ring elements individually specified to determine element-specific power Table 24: Tested Core Configurations (Vol in cm 3)
CORE No.
VOL SFE, No VOL SFE, VOL ALL FUEL SFE B Ring B Ring FFCR UNIT 1
201-206 4
49 46 15580.16 2337.02 17917.19 964.74 18881.93 53 50 17138.18 2337.02 19475.20 964.74 20439.94 57 54 18696.19 2337.02 21033.22 964.74 21997.96 59 56 19475.20 2337.02 21812.23 964.74 22776.97 65 62 21812.23 2337.02 24149.25 964.74 25113.99 73 70 24928.26 2337.02 27265.28 964.74 28230.02 74 71 25317.76 2337.02 27654.79 964.74 28619.53
Table 24: Tested Core Configurations (Vol in cm3)
CORE No.
VOL SFE, No VOL SFE, ALL SFE VOL ALL FUEL SFE B Ring B Ring FFCR 78 75 26875.78 2337.02 29212.80 964.74 30177.54 83 80 28823.30 2337.02 31160.32 964.74 32125.07 87 84 30381.32 2337.02 32718.34 964.74 33683.08 89 86 31160.32 2337.02 33497.35 964.74 34462.09 92 89 32328.84 2337.02 34665.86 964.74 35630.60 97 94 34276.36 2337.02 36613.38 964.74 37578.12 97 94 34276.36 2337.02 36613.38 964.74 37578.12 100 97 35444.87 2337.02 37781.89 964.74 38746.63 106 103 37781.89 2337.02 40118.92 964.74 41083.66 117 114 42066.44 2337.02 44403.46 964.74 45368.20 The number of fuel elements in Table 24 includes standard fuel elements (SFE) and fuel followers. Three fuel follower control rods at 321.58 cm 3 are included, and six individual B ring elements at 389.50 cm 3 each. Excess reactivity was derived from keff values calculated in KENO ("Transport k"). The number of fissions in regions within geometry units was taken from the KENO "Fission Densities" table.
5.1.1 Zero Power Critical Mass The result of the calculations for graphite and water-void configured cores is shown in Fig. 17 for room temperature (300 K) and nominal full power operation (600 K). Minimum fuel loading with graphite rods at 300 K is 60 elements (57 standard fuel elements and 3 fuel followers). Associated mass is 2.27 kg of 235U (slightly more than the 1.94 kg 231U in the GA TRIGA Mark reactor, slightly less than the 2.7 kg 235U in the GA Advanced TRIGA Prototype reactor, and comparable to 2.24 kg 235U of the GA TRIGA Mark Ill critical condition, op cit). The water-void configuration requires about 74 fuel element to achieve criticality at room temperature, 23% more fuel elements comparable to the 25% increase noted at GA.
Variation in Keff with Changes in Core Configuration
-H20 300 K -a-H20 600 K
-GR 300 K -GR 600 K 1.08 1.06 1.04 1.02 D 1.00 0.98 0.96 0.94 0.92 0.90 45 55 65 75 85 95 No. of Fuel Elements Figure 17: Excess Reactivity and Fuel Loading 105
5.1.2 Operational Critical Mass Compensation for elevated temperatures associated with power operation requires additional fuel to be loaded. Reactor operation at 600 K requires about 74 fuel elements with graphite rods in non-fuel spaces and K about 86 fuel elements in the water-void configured core. Table 25 shows the bounding range of possible core configurations considered (using fresh fuel) with excess reactivity high enough to support full power operations, and low enough to meet maximum requirements. Bounding values are provided for context with shading. Therefore analysis for the limiting core configuration considers the range of graphite-rod reflected cores greater with than 74 elements, and water-void configurations with greater than 86 fuel element elements.
Table 25: Excess Reactivity ($, All Rods Out) for Potential Core Configurations Number of Fuel Elements TEMP oK 73 74 78 83 87 89 92 97 100 Graphite Rod Configuration 300 350 450 550 600
$3.27
$4.11
$4.87
$5.69
$6.54
$6.85
$7.41
$3.05
$3.38
$4.44
$5.32
$5.98
$6.44
$7.18 Excess
$1.72
$2.32
$3.38
$4.12
$4.99
$5.63
$5.92 Reactivity
$0.3i
$1.04
$2.08
$2.75
$3.76
$4.04
$4.71 Too High
-$0.36
$0.11
$1.27
$2.17
$2.88
$3.21
$3.85 Water Void Configuration 300
$2.93
$4.19
$4.77
$5.48
$6.83
$7.44 350 Excess
$2.37
$4.04
$4.39
$4.95
$6.39
$7.00 450 Reactivity
$1.37
$2.81
$3.09
$4.11
$5.34
$5.89 550 Too Low
-$0.04
$1.23
$1.86
$2.73
$4.09
$4.43 600
-$0.86
$0.60
$1.15
$1.93
$3.38
$3.95 If the core configured with all graphite rods is loaded to 87 fresh fuel elements, excess reactivity at ambient temperature exceeds the limit. If the core configured with all water voids in non-fueled positions is loaded to 100 fresh fuel elements, excess reactivity at ambient temperature exceeds the limit. Introducing water voids in the graphite-configured core will reduce the core excess reactivity, and more fuel can be added as the core transitions from all graphite to all water voids. Depletion of fuel and the introduction of fission products associated with fission will reduce core excess reactivity and more fuel can be added as burnup increases. Therefore the operational core with fresh fuel considers graphite cores loaded with less than 89 fuel elements and water void configured cores with less than 97 fuel elements.
5.1.3 Fuel Element with Highest Power Density Elements in the B ring are exposed the highest neutron flux, and therefore generate the most power.
The ratio of the power in a fuel element to the average power per element in the core is the "Peaking Factor." The product of the average power and the peaking factor is used to calculate the power produced in a fuel element.
5.1.4 B RING PEAKING FACTORS KENO transport calculations provide fission density in each region of each geometry unit that contains fissionable material. Average fission density for the core is calculated as the sum of the total fissions for
all regions (the product of the fission density in a region and the volume of the region) normalized to the total volume of all regions with fissionable materials. The peaking factor of individually labeled units (i.e., all B ring positions) is calculated as the ratio of the fission density in the labeled unit to the (volume averaged) total fission density The maximum B ring peaking factor for each core configuration (that can support full power operation) at 300 K, 350 K, 450 K, 550 K, and 600 K are tabulated in Table 24. The peaking factors for the graphite configuration generally decreased slightly (on the order of 1%) as temperature increased from 300 K to 600 K. In all cases, the maximum peaking factor occurred in the B01 position. Cores configured with graphite elements have a more uniform power distribution and lower maximum peaking factors. The peaking factors in the graphite configurations decreased slightly (order of 1%) as temperature increased; this effect was not observed in the water-void core.
Table 26: Maximum B Ring Peaking Factors No.
GRAPHITE CONFIGURATIONS WATER VOID CONGIFURATION Elements 300 350 450 550 600 300 350 450 550 600 74 1.53 1.53 1.54 1.55 1.57 78 1.55 1.54 1.55 1.56 1.58 83 1.53 1.53 1.53 1.54 1.56 87 1.54 1.54 1.54 1.56 1.57 1.60 1.60 1.60 1.62 1.63 89 1.52 1.52 1.53 1.54 1.56 1.60 1.59 1.60 1.61 1.63 92 1.59 1.59 1.60 1.61 1.63 97 1.63 1.63 1.64 1.65 1.67 5.1.5 B RING ELEMENT POWER Average power per fuel element is the rated power distributed over all fuel elements. The maximum B ring fuel element power (Table 27) is product of the peaking factor from Table 25 and the average power for each core configuration. It is clear that the core configured with graphite has a higher power for equivalent reactivity values as fewer fuel elements are required with the graphite reflection. The maximum power occurs with 74 elements (73 elements will allow operation at less than full power and the data is provided only for context).
Table 27: Maximum B Ring Power (kW)
GRAPHITE CONFIGURATIONS WATER VOID CONGIFURATION No.
Max Power, B Ring Element Max Power, B Ring Element Elements 300 350 450 550 600 300 350 450 550 600 74 22.8 22.8 22.9 23.1 23.3 78 21.8 21.8 21.8 22.0 22.2 83 20.2 20.2 20.3 20.4 20.7 21.3 21.3 21.4 21.5 21.7 87 19.4 19.4 19.5 19.7 19.8 20.2 20.2 20.3 20.4 20.6 89 18.8 18.9 19.0 19.2 19.4 19.7 19.7 19.8 20.0 20.1 92 19.1 19.1 19.1 19.3 19.5 97 18.5 18.5 18.6 18.7 18.9
5.1.6 LIMITING CORE CONFIGURATION The B ring fuel element with the maximum power across all cores that that will support full power operation generates is 23.2 kW at 600 KW, 74 elements in the graphite core configuration.
5.2 Nuclear Characteristics of the Limiting Core Configuration KENO transport and ORIGEN buildup and decay calculations provide information representative of or useful in determining physics and operational parameters. Control rod positions are adjusted to achieve the desired conditions (critical or rods fully withdrawn, as applicable). The beginning of core life simulation assumes the limiting core configuration with unirradiated fuel then. The end of core life is simulation assumes a fully fueled core (with water void positions reserved for the neutron source and the pneumatic tube), operated at full power until criticality cannot be maintained at full power. Space is reserved in the core lattice for the neutron source and the pneumatic irradiation facility.
5.2.1 Physics Parameters and Flux Density Physics parameters and flux density are determined from simulation of the critical condition. All of the base units identified in Table 3 were used, with control rods in a banked position (equal Unit 4 and Unit 5 translations along the z axis). A series of calculations was performed for the LCC from 1 mW to full power at fuel temperature of 300 K with the control rods positioned at 18 cm withdrawn. Additional simulations were performed at full power operation, first with the same control rod position at temperature of 600 K and then with the control rod position adjusted to compensate for the elevated fuel temperature, approximately critical at control rod position 26.5 cm withdrawn. Table 28 values for "Neutron lifetime," "generation time," "mean free path" are taken from the KENO summary report.
Neutron flux from transport calculations at each power level is taken from the (energy) "group" report row totals, tabulated in Table 29.
Table 28: (Critical) Nuclear Physics Parameters Parameter 300 K 18cm 600 K 18 cm 600 K26 cm 1E-9 MW 1.1 MW 1.1MW 1.1 MW Neutron Lifetime 14.32 Its 14.32 Iis 14.77 I*s 14.62 ps Neutron Generation time 53.92 Its 53.92 Iis 54.83 IpS 54.45 ps Mean free path 1.011 cm 1.011 cm 1.003 cm 0.9964 cm nu bar 2.439 2.439 2.439 2.4385 During reactor operation, 235U is burned and fission products are generated, affecting reactor characteristics. Operation of a fully fueled core load was simulated as representative of changes to the operational core configuration over core life. At the end of core life with a keff of 1.005, neutron lifetime is 11.25 ps, neutron generation time 52.48 Ips, and system mean free path 1.017 cm.
Table 29: Flux Density & Fission, Absorption & Leakage Fractions and Power POWER Temp.
Rod keff thermal total fissions absorptions leakage Poiion flux flux 1.OOE-06 300 K 18 1.0037 1.07E+04 3.86E+04 1.00449 0.89233 0.107918 0.01 300 K 18 1.0037 1.07E+08 3.86E+08 1.00449 0.89233 0.107918 0.1 300 K 18 1.0037 1.07E+09 3.86E+09 1.00449 0.89233 0.107918 1 300 K 18 1.0037 1.07E+10 3.86E+10 1.00449 0.89233 0.107918 10 300 K 18 1.0037 1.07E+11 3.86E+11 1.00449 0.89233 0.107918 100 300 K 18 1.0037 1.07E+12 3.86E+12 1.00449 0.89233 0.107918 1000 300 K 18 1.0037 1.07E+13 3.86E+13 1.00449 0.89233 0.107918 1100 300 K 18 1.0037 1.18E+13 4.24E+13 1.00449 0.89233 0.107918 1100 600 K 18 0.9676 1.35E+13 4.53E+13 0.96865 0.890989 0.108998 1100 600 K 26 0.9982 1.35E+13 4.48E+13 0.99950 0.893075 0.107205 The flux density for the LCC core with fuel temperatures of 300 K is shown to be 1.19X10 14 n/cm2-s in the thermal range, and 4.30X101 4 n/cm2-s total, varying as 1.08X10 7 n/cm2-s per watt in the thermal range, and 3.91X10 7 n/cm 2-s per watt above thermal energy range. These values agree well with calculations reported by General Atomics12, where 2-D, 24 group calculations indicate average flux values for an 80 element 1 MW TRIGA to have 1.1X10 7 n/cm 2-s per Watt from 0 to 1 eV and 2.46X10 7 n/cm 2-s from 1 eV to 10 eV. At the end of core life as described, 1.1 MW power in the operational core requires a thermal neutron flux of 9.16X1012 n/cm 2-s, and total flux of 3.03X101 3 n/cm 2-s.
5.2.2 Element Peaking Factors As previously described, data from SCALE calculations to determine critical mass was used to evaluate maximum core peaking factors to support identification of the limiting core configuration, using the base geometry units as identified above. However, power within a fuel element is spatially distributed.
Therefore it was necessary to use the base geometry units of Table 3 and the optional unit 200, segmented first axially then radially. First, the unit 200 was segmented into 20 axial sections (Table 30),
then 21 radial segments (Table 31). The results of calculations using all base geometry units in Table 3 and alternate geometry unit 200 in position B01 are provided in Table 30 and Fig. 18 for axial distribution and Table 31 and Fig. 19 for radial distribution.
12 GA-4361, Calculated Fluxes and Cross Sections for TRIGA Reactors, G. B. West (August 14, 1963)
Table 30: B01 Axial Peaking Factor NODE 300 K 600 K 1
5.63E-01 5.64E-01 2
6.48E-01 6.40E-01 AXIAL PEAKING FACTOR 3
7.62E-01 7.24E-01 1.25-0 9.8 0 E -0.
4 8.59E-01 8.86E-01 CD 5
9.33E-01 9.80E-01 o
1.15 6
1.02E+00 1.01E+00 7
1.09E+00 1.12E+00 (D
1.05 8
1.12E+00 1.12E+00 9
1.2 0 E
- 0 0 1.1 6 E + 0 0 0.9........
10 1.19E+00 1.20E+00
.o 12 1.20E+00 1.17E+00 0.85 0.7 13 1.14E+00 1.16E+00 0
14 1.15E+00 1.13E+00 0,65 15 1.06E+00 1.07E+00 c
0.65 16 1.01E+00 9.66E-01 0
0.55 17 8.63E-01 9.09E-01
-20
-15
-40
-5 o
5 10 is 20 18 7.85E-01 7.60E-01 Axial Position (cm from centerline) 19 6.84E-01 6.65E-01 20 6.11E-01 6.04E-01 Figure 18, Axial Peaking Factor Table 31: Fuel Element B01 Radial Peaking Factor Radial Position Outer 1.826 1.753 1.680 1.606 1.533 1.459 1.386 1.313 1.239 1.166 1.092 1.019 0.946 0.872 0.799 0.725 0.652 0.579 0.505 0.432 0.358 Inner 1.753 1.680 1.606 1.533 1.459 1.386 1.313 1.239 1.166 1.092 1.019 0.946 0.872 0.799 0.725 0.652 0.579 0.505 0.432 0.358 0.285 Ave.
1.790 1.716 1.643 1.569 1.496 1.423 1.349 1.276 1.202 1.129 1.056 0.982 0.909 0.835 0.762 0.689 0.615 0.542 0.468 0.395 0.322 Temperature 300 600 1.972 2.088 1.853 1.918 1.718 1.732 1.583 1.599 1.481 1.484 1.380 1.365 1.284 1.271 1.183 1.168 1.096 1.067 1.022 0.989 0.933 0.910 0.849 0.828 0.779 0.763 0.713 0.701 0.649 0.634 0.574 0.570 0.513 0.503 0.441 0.437 0.387 0.383 0.324 0.323 0.267 0.266 Radial Peaking Factor 125 120..
115 S
.0.5-.-..-.-..---
0-0 OB05 SWK1
-6WK1
-i'..
UVIV
-*-FwI00
...~~
~ ~
f........
7-1 O0.00............ *"*....
i.1 0.0 02 0.4 0.6 0.z
- 1.
1.2 1.4 16 1.5 2.0 Radial Position (cm)
Figure 19, Element Radial Peaking factor
5.2.3 Burnup Effects Calculations were performed to determine the mass of uranium isotopes (235U and 238U) in a single fuel element (Fig. 20) for two conditions. The uranium burnup for the initial LCC core was determined, and the mass for a core with all locations loaded with fuel to predict the operational core at end of core life.
Excess reactivity was calculated from transport kef, and is shown in Fig. 21.
Uranium Burnup for LCC and Fully Loaded Core L(X-G-msU2M
-LM-.(C*
UZ
- --uI-r~~r:Gre-U238 r-F.lCo,:Gram.U235 100
- L00 o
- 4 0
"E E
2 MU 6 0 i....
40 20 0
200 400 600 800 1000 Burnup (MWD)
Figure 20: Uranium Burnup 1200 1400 Reactivity and Power History 14 12 -
10 6 B CU4...........i...
2 0
-2 0
100 200 300 400 500 600 700 Burnup (MWD)
Figure 21: Excess reactivity and Burnup 800 900 The change in excess reactivity following startup from a clean core to equilibrium full power operation at 300 K was simulated using the base geometry units in Table 3. Excess reactivity was determined from KENO Transport K values. Full power equilibrium excess reactivity decrease attributed principally to fission product poisons is shown in Fig. 22 to result in a reactivity deficit of approximately $3.5. Excess reactivity following shutdown from the operation was simulated, with the results provided in Fig. 23.
EXCESS REACTIVITY CHANGE FOLLOWING STARTUP WITH NO XENON
'U
'U O00
-0.5
-10
-1.5
-2.0
-2.5
-3.0
-3.5
-4.0
-4.5 10 20 30 40 50 Time afterStartup (h) 60 70 Figure 22, Excess Reactivity from Clean Core EXCESS REACTIVITY CHANGE FOLLOWING SHUTDOWN FROM EQ. XENON IP 2-3.5 3.0 2.5 2.
1.5 1.0 0.5 0.0
-0.5
-1.0
-1.3 0
1.2...
0 10 20 so 40 50 60 70 Time after Shutdown (h)
Figure 23: Excess Reactivity Following Shutdown Simulation of steady state reactor operations at 1.1 MW from initial operation of the LCC to the operational end of core life was performed to predict the behavior of fission product poisons and transuranic isotopes on system performance over time. Absorptions are used as a proxy for reactivity, but ORIGEN calculations are infinite medium, neglecting leakage. All base geometry units in Table 3 were used, with OPUS reports for neutron absorption for selected fission product poisons, uranium and transuranic isotope. Near the beginning of core life, 85.8% of all absorptions occur in the isotopes of interest, as listed in Table 32. A large fraction (on the order of 80-85%) of absorptions in these isotopes occur in 5 isotopes, U235, U238, Pu239, Xe135, and Sm151. The change in absorption for these isotopes over long term, steady state, full power operation is shown in Fig. 24.
Table 32: Isotope Absorption-Fractions ISOTOPE 0 MWD 66 MWD 198 MWD 264 MWD 396 MWD 462 MWD 550 MWD U235 8.02E-01 7.67E-01 7.46E-01 7.36E-01 7.14E-01 7.03E-01 6.88E-01 U238 5.93E-02 5.84E-02 6.04E-02 6.15E-02 6.38E-02 6.50E-02 6.68E-02 Pu238 2.46E-17 1.50E-08 3.66E-07 8.60E-07 2.95E-06 4.76E-06 8.25E-06 Pu239 6.79E-17 3.77E-03 1.15E-02 1.53E-02 2.27E-02 2.63E-02 3.09E-02 Pu240 5.46E-17 3.17E-05 2.97E-04 5.29E-04 1.19E-03 1.61E-03 2.28E-03 Pu241 7.54E-17 8.19E-07 2.35E-05 5.55E-05 1.85E-04 2.91E-04 4.86E-04 Pu242 7.03E-18 3.51E-10 3.24E-08 1.05E-07 5.55E-07 1.05E-06 2.16E-06 Xe131 8.24E-18 1.19E-04 4.25E-04 5.86E-04
,9.21E-04 1.10E-03 1.34E-03 Xe133 9.14E-18 4.79E-05 4.96E-05 5.05E-05 5.24E-05 5.34E-05 5.49E-05 Xe135 1.49E-13 2.41E-02 2.43E-02 2.43E-02 2.44E-02 2.44E-02 2.45E-02 Sm147 6.33E-18 1.14E-06 1.41E-05 2.60E-05 6.06E-05 8.33E-05 1.19E-04 Sm150 5.97E-18 2.75E-05 1.12E-04 1.57E-04 2.53E-04 3.03E-04 3.74E-04 Sm151 5.75E-16 1.15E-03 2.33E-03 2.60E-03 2.89E-03 2.96E-03 3.02E-03 Sm152 2.30E-17 5.08E-05 2.09E-04 3.06E-04 5.19E-04 6.33E-04 7.92E-04 Sm153 4.27E-17 2.06E-06 2.50E-06 2.76E-06 3.37E-06 3.72E-06 4.21E-06 TOTAL 86.1%
85.5%
84.6%
84.1%
83.1%
82.6%
81.9%
Major Absorption Fractions By Isotope u238 pu2S9 ne131
,--xe1S5 -
s,151
.-,.-u255 2 %
0 4%
C ~3%
0%
-11 80%
0 76%
74%
72% 0 "C,'
... 0 %
- 100 200 300 400 500 600 Burnuo (MWD)
Figure 24, Neutron Absorption in Major Isotopes 5.2.4 Fuel Temperature Reactivity Coefficient and Excess Reactivity As previously described, data from SCALE calculations using base units of Table 3 was performed to determine critical mass across a range of temperatures from 300 K to 600 K. The transport keff in each configuration and temperature variation was used to calculate net core reactivity, as reported in Table
- 33. The values for the LCC were plotted in Fig. 25 (along with other core configurations near the LCC).
The response for all data was remarkably similar, providing confidence that the temperature coefficient of reactivity can reliably be determined by the slope of the linear function relating temperature to reactivity change. The value of $-0.0127/A°K is approximately 11% lower than the value reported in the
original UT TRIGA SAR for a critical configuration of 64 elements. A similar fit for the 65 element core results in agreement with the original UT SAR to within about 6%.
Table 33: Excess Reactivity at Fuel temperature By Core Configuration:
Fuel Number of Fuel Elements Temp 65 73 74 78 83 87 89 92 97 100 106 (K)
Graphite Configuration 300
$1.61
$3.24
$4.19
$4.70
$5.64
$6.58
$6.83
$7.46
$8.24
$8.61
$9.55 300
$1.62
$3.29
$4.04
$5.04
$5.73
$6.50
$6.87
$7.36
$8.39
$8.71
$9.47 350
$1.16
$3.05
$3.38
$4.44
$5.32
$5.98
$6.44
$7.18
$8.02
$8.31
$8.97 450
-$0.03
$1.72
$2.32
$3.38
$4.12
$4.99
$5.63
$5.92
$6.80
$7.20
$8.09 550
-$1.65
$0.31
$1.04
$2.08
$2.75
$3.76
$4.04
$4.71
$5.45
$5.98
$6.71 600
-$2.29
-$0.36
$0.11
$1.27
$2.17
$2.88
$3.21
$3.85
$4.86
$5.07
$6.06 Water Void Configuration 300
-$2.81
-$0.21
$0.68
$1.94
$2.83
$4.13
$4.72
$5.49
$6.74
$7.47
$8.67 300
-$2.77
-$0.24
$0.77
$1.90
$3.03
$4.24
$4.82
$5.47
$6.92
$7.41
$8.64 350
-$3.53
-$0.83
$0.29
$1.34
$2.37
$4.04
$4.39
$4.95
$6.39
$7.00
$8.23 450
-$4.65
-$2.34
-$0.98
$0.30
$1.37
$2.81
$3.09
$4.11
$5.34
$5.89
$7.15 550
-$6.17
-$3.60
-$2.56
-$1.15
-$0.04
$1.23
$1.86
$2.73
$4.09
$4.43
$5.88 600
-$6.98
-$4.46
-$3.41
-$1.81
-$0.86
$0.60
$1.15
$1.93
$3.38
$3.95
$5.11 Evaluation of Fuel Temperature Reactivity Coefficient 65 El
-U-73 EL
-*74 El (LCC) -78 El.
-N-83 EL
$6JDO
$5.00
$4.00
$3.00
$2.00
$1.00 y, -.01~27x +
7.53 R'03935 T1....
ooo......................
$0-0
-$100 3..00 300 350 400 450 500 550 600 Fuel Temperature (K)
Figure 25: Excess Reactivity and Fuel Temperature 5.2.5 Excess reactivity & Shutdown Margin The base units in Table 3 and the fuel follower option Unit 8) were used in the Limiting Core Condition (i.e., 74 element core) and a fully filled core lattice (i.e., 117 fuel elements installed, water voids in space
reserved for source and pneumatic transit facility). For the LCC core, calculations were performed using material files generated following simulations of from 50 MWD, 100 MWD, 200 MWD and 300 MWD power histories. Calculations were performed to determine excess reactivity with the following conditions:
- 1) all control rods positioned at 38.1 cm;
- 2) all control rods positioned at 0 cm;
- 3) all control rods except the transient rod at 38.1 cm and the transient rod at 0 cm;
- 4) all control rods except shim rod 1 at 38.1 cm and shim rod I at 0 cm;
- 5) all control rods except shim rod 2 at 38.1 cm and shim rod 2 at 0 cm; Excess reactivity was determined directly from transport keff for the "all rods out" condition. The worth imposed by control rod insertion was calculated for the remaining conditions using the formula (where insertion is understood to cause negative reactivity worth):
= keff,ARO - keffRMn 1
$CR: kefARO-.kff CRi,,
0.007 The limiting shutdown margin (LSDM) is calculated as the shutdown margin with the most reactive control rod fully withdrawn. In all cases, the regulating rod was determined to be the most reactive control rod. The limiting shutdown margin was calculated in three separate ways. The label "LSDM 1" indicates a value based on keff from transport calculations configured with transient rod, shim 1 and shim 2 fully inserted, and the regulating rod fully withdrawn. This directly represents the limiting control rod configuration and is therefore has the most confidence. Values labeled "SDM" were calculated by assuming the reactivity worth of all rods out, decreased by the reactivity worth of inserting the transient rod, shim 1, and shim 2. Since control rod worths are interdependent, this is less representative of the core configuration but closely matches experimental determination of the limiting shutdown margin. The results are provided in Table 36.
Table 36: LCC Control Rod and Excess Reactivity ($) and Burnup MWD ARO ARI RR TR SH1 SH2 ROD-SUM 8.3 6.94 16.90 4.36 3.05 2.97 2.80 13.18 10 6.71 16.72 4.10 3.31 2.63 2.64 12.68 15 6.81 17.05 4.57 3.28 3.08 3.27 14.21 20 6.32 17.05 4.37 3.37 3.10 3.12 13.95 30 6.05 17.05 4.37 3.37 3.10 3.12 13.95 100 4.05 17.20 4.20 3.05 2.68 10.95 20.87 LSDM 3.82 3.75 4.21 4.67 4.67 6.90 SDM 1.87 1.88 2.83 3.27 3.53 12.62 ARO-ARI 9.96 10.01 10.24 10.73 11.00 13.15
5.2.6 Burnup effects Over time, the limiting core configuration will require increasing the number of fuel elements to compensate for burnup until all locations reserved for fuel are filled. Therefore, burnup calculations are based on a full core load. A simulation of steady state reactor operations at 1.1 MW from initial operation to end of core life provides data on flux density, changes in uranium mass, and the effects of significant isotopes generated during operation over core life. Thermal and total neutron flux in each material is calculated by KENO; the average flux values for the ZrH fuel material shown in Fig. 26. The total 235U and 23U mass at each burnup interval is calculated by ORIGEN, and reported by OPUS as indicated in Fig. 27. Excess reactivity derived from transport calculations over core life are provided in Fig. 28. Neutron absorption in 235U and 238U (as a fraction of total absorptions) is shown in Fig. 29. The absorption fractions for other significant isotopes generated during operations are shown in Fig. 30.
Average Neutron Flux in Fuel, 117 Element Core 41+13 3*E+13
" 31+13
.21+13 X
LL 21+13 0
11+13 Z
51+12 01400 Thermal FlUX
'~-Total FlUX 0
100 200 300 400 500 600 700 80 Burnup (MWD)
Figure 26: Neutron Flux in Fuel of a 117 Element Core as a Function of Burnup Mass of z*U Versus Burnup Ln 0n 45SE+0 4.4E403S 4.3M+03 4 2E+03 5.8E+03 3.7E+03 3.6E4-03 3.5E+03 3.2E+403 32IE+03 SýX+03 0
100 200 300 400 500 600 700 Burnup (MWD) 000 900 low0 Figure 27: Uranium (235 and 238) Mass in a 117 Element Core as a Function of Burnup
Excess reactivity versus Bumup (Full Core Load)
$14"-
114 10 I
-$04
+ $12.14 1
$4 1-04 1.02 0
100 200 300 400 500 600 700 800 Burnup (MWD)
Figure 28: Excess Reactivity in a 117 Element Core as a Function of Burnup Absorption Fractions Versus Burnup
-*-u238 -u235 6.4%
82%
6.3% A 804% -n
- 56%
.r 170%
041
.5%
5.4%
,768%
0200 400 600 800 Figure 29: Fraction of Neutrons Absorbed in 235u and 238U (117 Element Core) vs. Burnup
Absorption Fractions Versus Bumup for Selected Isotopes w29
xel5 -sm151 u235 2
-4
-4
- VCI0a, x
L')
CV)
-4.
0~
C C
E U-8^%
78%
7216 70%
68%
0 100 200 O
400 500 6W 700 O
Figure 30: Fraction of Neutrons Absorbed in Selected Isotopes (117 Element Core) vs. Burnup 5.2.7 Experiment effects As shown in Table 36, actual shutdown margin based on ke as determined with the most reactive rod fully withdrawn with fresh fuel is shown to be $3.82. However, the method of calculating excess reactivity consistently underestimates excess reactivity by a factor of two. Therefore the limit on experiment worth provides assurance that excess reactivity limits with the most reactive shutdown margin fully with draw is met.
5.2.8 Accident Source terms A simulation was performed with a fully loaded core (all core spaces filled with fresh standard fuel elements) operated at 1.1 MW until kexcess was less than unity. The simulation then decayed the core for 20 minutes, simulating the amount of time after shutdown required to remove a fuel element from the core. Activity of the major halogens and iodine isotopes was calculated for strategic time intervals, and is provide in Table 37. Total decay heat was calculated for time intervals after shutdown (Fig. 30).
Table 37: Fission product Inventory, Maximum Single Fuel Element TIME 20M [50 M 7.5 H 11.5 H 12 H 1 D 7 D 30 D 180 D 365 D br82 br83 br84m br84 br85 br86 br87 i131 i132 i133 i134 i135 1.82E-1 6.37E1 2.25 1.12E2 1.61E2 1.98E2 2.39E2 3.56E2 5.34E2 8.05E2 9.40E2 7.60E2 1.81E-1 1.80E-1 1.57E-1 1.45E-1 1.14E-1 6.75E-3 1.32E-7 1.64E-11 1.63E-11 6.17E1 5.61E1 7.75 2.44 7.61E-2 5.80E-9 5.80E-9 5.80E-9 5.80E-9 2.17E-1 6.67E-3 2.05E-10 2.05E-10 2.05E-10 2.05E-10 2.05E-10 7.93E1 4.12E1 4.44E-3 2.35E-5 1.02E-8 1.02E-8 1.02E-8 1.51 1.12E-3 1.46E-8 1.47E-8 1.47E-8 1.47E-8 1.47E-8 2.05E-10 1.02E-8 1.47E-8 5.78E-5 1.80E-8 1.80E-8 1.80E-8 1.80E-8 1.80E-8 1.80E-8 1.80E-8 6.58E-5 2.17E-8 2.17E-8 2.17E-8 2.18E-8 2.18E-8 2.18E-8 2.18E-8 3.56E2 3.55E2 3.48E2 3.43E2 3.30E2 1.98E2 2.72E1 6.38E-5 5.34E2 5.32E2 5.04E2 4.86E2 4.37E2 1.19E2 8.23E-1 4.82E-8 2.05E-10 1.02E-8 1.47E-8 1.80E-8 2.18E-8 3.24E-8 4.82E-8 8.02E2 7.94E2 6.37E2 5.57E2 3.74E2 8.85E2 7.44E2 7.12 3.28E-1 2.62E-5 3.08 1.05E-7 7.37E-8 7.37E-8 8.57E-8 8.57E-8 8.57E-8 8.57E-8 7.33E2 6.96E2 3.33E2 2.18E2 6.15E1 1.56E-5 6.92E-8 6.92E-8 6.92E-8
Table 37: Fission product Inventory, Maximum Single Fuel Element TIME 20 M so M 7.5 H 11.5 H 12 H 1 D 7 D 30 D 180 D 365 D i136 3.15E2 1.46E-2 3.31E-8 2.87E-8 2.87E-8 2.87E-8 2.87E-8 2.87E-8 2.87E-8 2.87E-8 kr83m 6.28E1 6.27E1 6.19E1 1.94E1 7.30 2.86E-1 2.67E-7 2.23E-7 7.07E-8 2.04E-8 kr85m 1.55E2 1.49E2 1.38E2 4.69E1 2.52E1 3.94 1.50E-8 1.42E-8 1.42E-8 1.42E-8 kr85 9.59 9.59 9.59 9.59 9.59 9.59 9,58 9.54 9.29 9.00 kr87 3.00E2 2.52E2 1.92E2 4.26 4.79E-1 6.91E-4 2.73E-8 2.73E-8 2.73E-8 2.73E-8 kr88 4.06E2 3.74E2 3.31E2 6.01E1 2.26E1 1.21 3.70E-8 3.70E-8 3.70E-8 3.70E-8 kr89 5.18E2 6.03 7.96E-3 4.72E-8 4.72E-8 4.72E-8 4.72E-8 4.72E-8 4.72E-8 4.73E-8 xe131m 4.25 4.25 4.25 4.25 4.25 4.24 3.87 1.61 3.49E-4 7.23E-9 xe133m 8.93 8.93 8.93 8.79 8.66 8.09 1.57 1.10E-3 8.16E-10 8.16E-10 xe135 3.55E2 3.65E2 3.78E2 4.19E2 3.80E2 2.20E2 6.79E-3 7.26E-8 7.26E-8 7.26E-8 xe137 7.38E2 1.97E1 8.28E-2 6.71E-8 6.71E-8 6.72E-8 6.72E-8 6.72E-8 6.72E-8 6.72E-8 xe138 7.50E2 2.77E2 6.28E1 1.37E-7 6.82E-8 6.83E-8 6.83E-8 6.83E-8 6.83E-8 6.83E-8 Decay Heat from Single. High Power Fuel Element
=I' LOW -
900 L
800 700 600 500 400 30O 200 1001 0.01 0.1 1
10 100 1000 Time After Shutdown (Days)
Figure 30, Decay Heat Following Steady State Full Power, End of Life Operations 6.0 Model Validation Validity of the SCALE model application to the UT TRIGA reactor is verified in two ways. First, the prediction of mass required to support operations is compared to the 1992 operational core loading.
Second, data from model calculations is used to evaluate core excess reactivity and the reactivity worth of individual control rods with values compared to measured data for the 1992 operational core.
6.1 Model Comparison with Historical Reactors As previously shown (Fig. 7), the SCALE model predicts criticality with fresh fuel and graphite moderation to be 60 elements, and 86 for water moderation. A fully operational core was assembled which included 84 lightly burned standard fuel elements and three fresh fuel followers augmented with 18 graphite rods, leaving 13 water voids in the G ring. According to the figure, SCALE predicts excess reactivity (for 87 fuel elements, 84 standard and 3 followers) of approximately $1.00 with fresh fuel and
a water-void configured core, $7.00 with a full graphite-configured core. The lightly burned fuel with a measured excess reactivity of approximately $6.4 is within the predicted range for fresh fuel.
6.2 Reactivity Values Data from calculations using the SCALE model of the UT TRIGA reactor was used to determine excess reactivity and the total reactivity worth of each control rod. Excess reactivity was calculated based on keffwith the control rods in a fully withdrawn position (keff.ARO). The worth of an individual control rod
($cR) was based on the previous calculation and keffcalculation with the control rods in a fully inserted position (keffCRi) as follows:
keff'ARO -
keffCR,M 1
$CR -kffARO
- erCR, 0.007 Reactivity values calculated from data generated using the SCALE model for control rod worth and excess reactivity measurements are compared to measurements accomplished in 1992. The 1992 core 235U mass and enrichment of the 1992 are used to adjust material composition in the SCALE model to reflect conditions representative of the fuel loaded in the 1992 core. All of the standard fuel elements used in 1992 had prior power history, lightly burned at the previous UT TRIGA reactor at Taylor Hall or at General Atomics facilities. Some of the elements had a power history at GA facilities followed by operation at the UT Taylor Hall facility. All of the fuel elements decayed approximately one year following removal prior to installation at the current UT TRIGA located at the NETL.
Although records indicate significant variation for fuel burnup, 30 of the inner ring elements had power history of approximately 0.243 MWD per element from operation at the Taylor Hall reactor on the UT campus. The average burnup had a standard deviation of 0.035 with a maximum of 0.304 MWD and a minimum of 0.144 MWD at Taylor Hall; a single element in this group had 0.657 MWD burnup from a GA facility (C11 position). Fifty-four of the outer elements were documented to have burn of approximately 0.676 MWD per element at GA facilities, with one exception, a single element with 30% higher burnup.
The average burnup of this group had a standard deviation of 0.025 MWD with a maximum value of 0.752 MWD and a minimum of 0.619 MWD.
As indicated in Fig. 3, two of the elements with the higher average burnup value are located in a ring with the inner set of elements with lower burnup for the remaining standard fuel elements; to simplify the description, all elements with the lowest average power history will be referred to as "inner elements" and elements with the higher power history will be referred to as "outer elements."
A series of calculations was accomplished to determine isotopic concentrations of fuel at various burnup intervals. The power history for the previous UT TRIGA is complex. The UT records did not indicate variation expected from operation in various rings of the Taylor Hall circular TRIGA core, i.e. based on different core-radial peaking factors. Facility records report all of the Taylor Hall fuel to have initially contained 38 grams of 235U; the lack of variation in assay values is unusual. UT does not have records relating the operation of the individual fuel elements to specific GA reactors, which do not all have the same rated power. The UT records do not contain information about the power level for the burnup of the GA fuel. Assumptions required to accurately model fuel composition based on prior power history are therefore extremely challenging.
Lacking information that would fully characterize the power history of the lightly burned full in the 1992 UT TRIGA core, all previous operation is assumed to have occurred at 250 kW. Simulations were made assuming 250 kW fresh TRIGA fuel in the 1992 core configuration, operated to target burnup values. The fresh fuel composition was based on the average value of 235U mass and enrichment from special nuclear material inventory records. Target burnup value was based on total power generation for all of the lightly burned elements. Initial attempts to model the lightly burned fuel composition resulted in calculated excess reactivity much greater than excess reactivity measured in 1992.
Since the power history for 30 elements in the inner rings was approximately 1/3 of the power history for the outer elements and fuel followers were fresh fuel in high reactivity worth positions, three different material specifications were developed to simulate the core composition more accurately.
One material specification was used to define fresh fuel, one to simulate material with average burnup for the inner elements, and one to simulate material with average burnup for the outer elements. While the results showed excess reactivity much closer to the measured 1992 values, excess reactivity calculations still greatly exceeded measured values.
Based on uncertainty in characterization of power history, previously developed material specifications were used in a set of calculations with the outer element burn three times the value of the inner element burn. With inner element material specification based on an average burnup of 0.69 MWD per element, the reactivity values calculated from the model show general agreement with measured values.
Reactivity values calculated from SCALE data and two sets of reactivity measurements from 1992 (immediately following core loading on 3/25/1992, and at the concluding of physics testing on 7/23/1992) are provided in Table 38. The column A$ref is the difference between the measured 1992 value and the value calculated from SCALE data. The A$ref/$ column is the ratio of the difference to the reference 1992 value (expressed as a per cent).
Table 38: 1992 UT TRIGA REACTIVITY ($) SCALE, SURV-6 & SURV 3 PARAMETER SCALE 03/25/1992 A~$ef A$ref/$
07/23/1992 A$ref A$ref/$
EXCESS
$6.76 NA NA NA
$6.38
$0.38 0.2%
REG ROD
$4.25
$4.59
$0.34
-7.5%
$4.08
$0.17 4.2%
TR ROD
$3.01
$3.34
$0.53
-15.1%
$3.26
-$0.26
-7.8%
SHIM 2
$3.26
$3.46
$0.20
-5.6%
$3.30
-$0.04
-1.2%
SHIM 1
$3.02
$3.32
$0.30
-8.9%
$3.17
-$0.15
-4.8%
SUM RODS
$13.53
$14.90
$1.37 9.2%
$13.82
-$0.28 0.04 6.3 Conclusion Core loading for the 1992 fuel using a fraction of non-fuel spaces using graphite rods is consistent with calculations of loading to support full power operation. Flux density calculated for the LCC (74 element, graphite rod configured) core agrees with historical calculations performed by General Atomics. There is good agreement between measured and calculated reactivity worth values when nominal material specification is adjusted to bring excess reactivity into agreement. Therefore, results of calculations using the SCALE UT TRIGA model provide confidence that the model is capable of adequately predicting reactor performance.
ATTACHMENT 2: UT TRIGA Historical Core Data Table 1 provides information about the UT TRGIA core in 1992. Initial critical was achieved with standard fuel elements in positions F06 and F1l, but these elements were removed to reduce excess reactivity on 3/19/1992 prior to completion of physics testing. In the TYPE column: SFE indicates standard fuel element, IFE indicates instrumented fuel element, and FFCR indicates fuel follower. The initial 23SU mass (g) is provided as INIT. 235, and the mass in the element at initial criticality is 1992 235.
Most of the fuel had prior power history, represented as GA BRN for fuel lightly irradiated at GA reactors and TH BRN for power history at the Taylor Hall TRIGA reactor located on the UT main campus.
Table 1: UT TRIGA 1992 Core Inventory (Physics Testing)
INIT.
1992 GA TH INIT.
1992 GA TH 235 235 BRN BRN 235 235 BURN BURN BO0 B02 B03 B04 B05 B06 C0l C02 C03 C04 C05 C06 C07 C08 C09 Cdo Cli C12 D01 D02 D03 D04 D05 D06 D07 D08 D09 D10 Dl1 D12 D13 D14 D15 5921 SFE 5922 SFE 5981 IFE 6143 SFE 6886 SFE 6889 SFE 5916 SFE 5917 SFE 6924 SFE 6926 SFE 6927 SFE 10148 FFCR 6928 SFE 6929 SFE 6930 SFE 5283 IFE 6932 SFE 5844 SFE 5845 SFE 5846 SFE 5902 SFE 5903 SFE 10146 FFCR 5904 SFE 5912 SFE 5913 SFE 5914 SFE 5915 SFE 5918 SFE 5919 SFE 10147 FFCR 5920 SFE 38.00 37.76 38.00 37.76 38.00 37.71 38.00 37.85 38.00 37.76 38.00 37.76 38.00 37.78 38.00 37.80 38.00 37.76 38.00 37.76 38.00 37.76 32.26 32.26 38.00 37.76 38.00 37.76 38.00 37.76 38.00 37.31 38.00 37.76 38.00 37.68 38.00 37.68 38.00 37.68 38.00 37.69 38.00 37.69 31.57 31.57 38.00 37.70 38.00 37.76 38.00 37.76 38.00 37.76 38.00 37.76 38.00 37.76 38.00 37.76 31.63 31.63 38.00 37.76 I
0.657 0.233 0.225 0.277 0.144 0.233 0.233 0.207 0.190 0.233 0.233 0.233 0.233 0.233 0.233 0.233 0.304 0.304 0.304 0.296 0.296 0.283 0.233 0.233 0.233 0.233 0.233 0.233 0.225 El0 Ell E12 E13 E14 E15 E16 E17 E18 E19 E20 E21 E22 E23 E24 FO0 F02 F03 F04 F05 F06 F07 F08 F09 F10 Fll F12 F13 F14 F15 F16 F17 F18 2912 2913 2915 2918 2925 2927 2928 2929 2930 2932 2935 2938 2939 2940 2941 2944 2946 2947 2948 2950 2952 2954 2955 2957 2959 2960 2962 2964 2965 2968 2969 SFE SFE SFE SFE SFE SFE SFE SFE SFE SFE SFE SFE SFE SFE SFE SFE SFE SFE SFE SFE GR SFE SFE SFE SFE GR SFE SFE SFE SFE SFE SFE SFE 38.95 38.24 40.11 39.38 39.04 38.33 38.68 37.97 36.76 36.07 40.23 39.50 37.68 36.99 38.35 37.65 39.78 39.05 37.76 37.07 40.03 39.30 40.11 39.38 37.40 36.72 39.41 38.69 36.78 36.09 36.58 35.89 38.79 38.08 39.37. 38.65 40.57 39.83 38.34 37.64 42.43 41.66 40.04 39.31 38.40 37.68 40.17 39.44 37.38 36.70 39.59 38.85 37.64 36.95 37.22 36.54 36.34 35.66 40.02 39.29 39.11 38.40 0.676 0.695 0.676 0.676 0.638 0.695 0.657 0.667 0.695 0.657 0.695 0.695 0.648 0.686 0.638 0.638 0.676 0.686 0.705 0.667 0.733 0.695 0.667 0.695 0.648 0.686 0.657 0.648 0.629 0.695 0.676 0.014 0.014 0.014 0.014 0.014 0.014
Table 1: UT TRIGA 1992 Core Inventory (Physics Testing)
ID TYPE NIT.
1992 GA TH INIT.
1992 GA TH 235 235 BRN BRN 235 235 BURN BURN D16 6142 SFE D17 6923 SFE D18 6925 SFE E01 2899 SFE E02 2902 SFE E03 2903 SFE E04 2904 SFE E05 2905 SFE E06 2906 SFE E07 2908 SFE E08 2910 SFE E09 2911 SFE 38.00 37.71 0.277 38.00 37.76 0.233 38.00 37.76 0.233 36.17 35.49 0.629 0.014 39.04 38.33 0.676 40.05 39.30 0.695 0.014 42.72 41.94 0.743 38.15 37.43 0.667 0.014 37.72 37.03 0.657 40.30 39.56 0.705 37.27 36.59 0.648 38.73 38.00 0.676 0.014 F19 2970 SFE 40.23 39.50 0.695 F20 2971 SFE 39.58 38.86 0.686 F21 2974 SFE 38.04 37.35 0.657 F22 2975 SFE 38.50 37.78 0.667 0.014 F23 2976 SFE 40.41 39.67 0.705 F24 2977 SFE 37.82 37.13 0.657 F25 2979 SFE 36.91 36.22 0.638 0.014 F26 2983 SFE 36.72 36.03 0.638 0.014 F27 2894 SFE 42.92 42.87 0.051 F28 2985 SFE 35.74 35.07 0.619 0.014 F29 5198 SFE 40.00 35.07 4.695 F30 3513 SFE 39.00 38.01 0.752 0.190 The core inventory as indicated in Table 2 in 2012 is taken from special nuclear material records, with the power history indicated as MWD.
Table 2: Core Inventory, 2012 POS ID TYPE MWD POS ID TYPE MWD POS ID TYPE MWD BO1 B02 B03 B04 B05 B06 Col C02 C03 C04 C05 C06 C07 C08 C09 c10 Cll C12 D01 D02 D03 D04 D05 D06 D07 D08 2985 3384 10878 3013 2899 10708 2965 2984 2944 2931 2983 10148 2980 2925 2941 2979 2964 2910 2959 2906 2992 2962 10146 2928 2939 SFE SFE IFE SFE SFE IFE TR SFE SFE SFE SFE SFE FFCR SFE SFE SFE SFE SFE SFE SFE SFE SFE SFE FFCR SFE SFE 3.295 2.130 2.297 3.532 3.304 2.490 3.304 3.290 3.314 2.837 3.314 2.661 2.924 3.314 3.314 3.314 3.309 3.309 3.309 3.318 3.742 3.318 2.661 3.318 3.309 E05 6886 E06 5912 E07 5846 E08 5903 E09 5917 El0 6929 Ell E12 6925 E13 5844 E14 6923 E15 5919 E16 5921 E17 6927 E18 5902 E19 5904 E20 6930 E21 6889 E22 5914 E23 6142 E24 6928 F01 10817 F02 5911 F03 3496 F04 3504 F05 3703 F06 10816 SFE SFE SFE SFE SFE SFE 3EL SFE SFE SFE SFE SFE SFE SFE SFE SFE SFE SFE SFE SFE SFE SFE SFE SFE SFE SFE 2.894 2.894 2.965 2.957 2.851 2.894 2.894 2.965 2.894 2.894 2.894 2.894 0.296 2.944 2.894 2.894 2.894 2.938 2.894 1.854 2.439 2.681 2.671 2.681 1.854 F21 2971 SFE 3.347 F22 2969 SFE 3.337 F23 6926 SFE 2.894 F24 3513 SFE 3.603 F25 10811 SFE 1.854 F26 2960 SFE 3.361 F27 2947 SFE 3.347 F28 2911 SFE 3.352 F29 5922 SFE 2.886 F30 10814 SFE 1.854 G02 10704 SFE 1.475 G03 2908 SFE 3.366 G04 3700 SFE 2.633 G05 6931 SFE 0.698 G06 5920 SFE 2.886 G08 10701 SFE 2.213 G09 2957 SFE 3.356 G10 2938 SFE 3.189 Gl1 2927 SFE 3.356 G12 10702 SFE 2.308 G14 2970 SFE 3.356 G15 2976 SFE 3.366 G16 2952 SFE 2.413 G17 10815 SFE 1.854 G18 2904 SFE 3.404 G20 2968 SFE 3.356
Table 2: Core Inventory, 2012 POS ID TYPE MWD POS ID TYPE MWD POS ID TYPE MWD D09 5918 D10 2977 Dll 2974 D12 2905 D13 2943 D14 10147 D15 2950 D16 2929 D17 2955 D18 2975 E01 5845 E02 6932 E03 2932 E04 5915 SFE SFE SFE SFE SFE FFCR SFE SFE SFE SFE SFE SFE SFE SFE 2.894 3.318 3.318 3.342 2.880 2.661 3.328 3.328 3.342 3.342 2.965 2.894 3.318 2.894 F07 2915 F08 2946 F09 6924 F10 10812 Fll 2958 F12 5913 F13 F14 F15 2902 F16 10813 F17 2912 F18 6143 F19 5916 F20 2940 SFE SFE SFE SFE SFE SFE 3EL 3EL SFE SFE SFE SFE SFE SFE 3.337 3.337 2.894 1.854 3.038 2.894 3.337 1.854 3.337 2.805 2.868 3.347 G21 2903 SFE 3.518 G22 2935 SFE 3.356 G23 2930 SFE 3.356 G24 2951 SFE 3.213 G26 10699 SFE 1.475 G27 2948 SFE 3.366 G28 2913 SFE 3.356 G29 2954 SFE 3.356 G30 10700 SFE 2.427 G32 SRC G33 2918 SFE 3.337 G34 PNT G35 10810 SFE 1.854 G36 10703 SFE 1.475