ML15091A749
| ML15091A749 | |
| Person / Time | |
|---|---|
| Site: | SHINE Medical Technologies |
| Issue date: | 04/01/2015 |
| From: | Bynum R V SHINE Medical Technologies |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| SMT-2015-011 NSA-DAC-SHN-13-02, Rev. 1 | |
| Download: ML15091A749 (15) | |
Text
2555 Industrial Drive l Monona, WI 53 713 l P (608) 210-1060 l F (608) 210-2504 l www.shinemed.com April 1 , 201 5 SMT-201 5-011 10 CFR 50.30 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
References:
(1) SHINE Medical Technologies, Inc. letter to NRC, dated March 26, 2013, Part One of the SHINE Medical Technologies, Inc. Application for Construction Permit (ML130880226)
(2) SHINE Medical Technologies, Inc. letter to NRC, dated May 31, 2013, Part Two of the SHINE Medical Technologies, Inc. Application for Construction Permit (ML13172A324
) (3) NRC letter to SHINE Medical Technologies, Inc., dated September 19, 2014, SHINE Medical Technologies, Inc. - Request for Additional Information Regarding Application for Construction Permit (TAC Nos. MF2305, MF2307, and MF2308) (ML14195A159)
(4) SHINE Medical Technologies, Inc. letter to NRC, dated October 15, 2014 , SHINE Medical Technologies, Inc. Application for Construction Permit , Response to Request for Additional Information (ML14296A190
) (5) SHINE Medical Technologies, Inc. letter to NRC, dated December 3, 2014 , SHINE Medical Technologies, Inc. Application for Construction Permit , Response to Request for Additional Information (ML14356A528
) SHINE Medical Technologies, Inc. Application for Construction Permit Submittal of NSA-DAC-SHN-13-02, Revision 1 Pursuant to 10 CFR 50.30, SHINE Medical Technologies, Inc. (SHINE) submitted an application for a construction permit to construct a medical isotope facility to be located in Janesville, WI (Reference s 1 and 2). Via Reference (3), the NRC staff determined that additional information was required to enable the staff's continued review of the SHINE construction permit application.
SHINE responded to the NRC staff's requests for additional information (RAIs) via Reference s (4) and (5). The SHINE Response to RAI 6b.3-1, provided via Reference (5), provided NSA-TR-07-08, R e vision 0, as the SHINE project-specific validation report.
However, SHINE should have provided NSA-DAC-SHN-13-02, Revision 1 , as the SHINE project-specific validation report. Enclosure 1 provides NSA-DAC-SHN-13-02, Revision 1 , supporting the SHINE Response to RAI 6b.3-1. If you have any questions, please contact Mr. Jim Costedio, Licensing Manager, at 608/210-1730.
Document Control Desk Page 2 I declare under the penalty of perjury that the foregoing is true and correct. Executed on April 1, 2015. Very truly yours, R. Vann Bynum, Ph.D. Chief Operating Officer SHINE Medical Technologies, Inc. Docket No. 50-608 Enclosure cc: Administrator, Region Ill, USNRC Project Manager, USNRC Environmental Project Manager, USNRC Supervisor, Radioactive Materials Program, Wisconsin Division of Public Health 12 pages follow ENCLOSURE 1 SHINE MEDICAL TECHNOLOGIES, INC.
SHINE MEDICAL TECHNOLOGIES, INC. APPLICATION FOR CONSTRUCTION PERMIT SUBMITTAL OF NSA
-DAC-SHN-13-02, REVISION 1 NSA-DAC-SHN-13-02, REVISION 1 EVALUATION OF BIAS OF URANIUM SULFATE SOLUTIONS FOR MCNP 5 AND THE ENDF/B
-VI CROSS SECTION LIBRARY NSA-DAC-SHN-13-02 Rev. 1 Page 2 of 13 Revision History Log Rev Date Revision Description/Reason 0 Original Issue 1 Editorial changes and clarifications based on AECOM review.
- Defined ZAID in Section 1.2.
- Defined ENDF/B
-VI in Section 5.0.
- Clarified experiment report number from Reference 1.
NSA-DAC-SHN-13-02 Rev. 1 Page 3 of 13 Table of Contents
1.0 INTRODUCTION
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1.1 BACKGROUND
/PURPOSE ................................
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.. 4 1.2 LIMITS OF APPLICABILITY
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2.0 CONCLUSION
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3.0 REFERENCES
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5 4.0 ANALYSIS/PROCESS METHODOLOGY ................................
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.... 5 5.0 COMPUTER CODES USED IN DAC ................................
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5 6.0 ASSUMPTIONS & OPEN ITEMS ................................
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7 6.1 ASSUMPTIONS
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7 6.2 OPEN ITEMS ................................
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7 7.0 ACCEPTANCE CRITERIA
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7 8.0 CALCULATIONS
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7 8.1 METHOD DISCUSSION ................................
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...... 7 8.2 INPUT ................................
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8 8.3 EVALUATIONS
, ANALYSIS , AND DETAILED CALCULATIONS
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8 APPENDIX 1: REPRESENTATIVE INPUT FILES
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11 NSA-DAC-SHN-13-02 Rev. 1 Page 4 of 13
1.0 INTRODUCTION
1.1 Background/Purpose Nuclear criticality safety analysis is performed for all systems and operations involving the handling, storage and processing of fissile material
. The nuclear criticality safety analysis establishes the nuclear criticality safety operating limits for the systems and operations. Calculation methods are used to provide an estimate of critical conditions and the margin o f
subcriticality for the systems and operations under evaluation. The computational methods predict the neutronic behavior of the system and operation. However, certain approximations are inherent in the computer code used including inexact neutron cross section data and statistical uncertainty.
Validation compares the computational method with documented critical experiments to determine any bias that might exist between the calculated reactivity of a given system and the actual conditions. It is a process that determines and establishes computational method applicability, adequacy, and uncertainty.
This report evaluates the bias associated with uranium sulfate critical experiments using the MCNP 5 computer code system. The critical experiments are modeled as reported in NEA/NCS/DOC (95)03 (Reference 1, experiment HEU-SOL-THERM-046). This report and the evaluated bias are used in conjunction with the bias evaluation for MCNP 5 in Reference s 2 and 3.
1.2 Limits
of Applicability
While the experiments used herein are from a limited area of applicability, the results are judged to be applicable to all thermal neutron systems evaluated with MCNP 5. Applications using the conclusions established for this experiment data set must use the isotopic material representations of Reference 2 and Table 2 to ensure that the default cross section library is used.
2.0 CONCLUSION
S Uranium sulfate solution benchmark critical cases are modeled using MCNP 5 with the default cross section library
. This report documents the methodology and results.
The calculation results for the critical experiments chosen for code bias evaluation are summarized in Table 1.
The experiments include high enriched uranium (HEU) in the thermal energy range. Though the experiment cases here are all with highly enriched uranium, this study provides evidence that the uranium complexed with sulfate solution does not introduce a significant, non
-conservative bias to the calculation method compared to the benchmark studies run for other complexes at high enrichment. The code treatment of the cross sections for other materials is well
-documented and demonstrated at the full range of enrichment, and is shown in Reference 2 to give a slight positive bias as do the sulfate cases; therefore, it is reasonable to conclude that the use of the sulfate complex with lower enrichment uranium will not exhibit significant, non
-conservative bias. The behavior of all the solution complex materials is dominated by the neutron moderating NSA-DAC-SHN-13-02 Rev. 1 Page 5 of 13 materials present, and the presence of the sulfur does not significantly affect the neutron spectrum or behavior.
It is concluded that the MCNP 5 calculated keff data with the use of uranium sulfate should use the same bias as with the use of uranium homogeneous systems for thermal neutron energies. These conclusions are valid for thermal systems only.
Table 1: HEU-SOL-THERM-046 Benchmark Critical Experiment Results Case 235 U Enrichment Uranium Form Geometry Moderator / Reflector H/235 U Average Neutron Energy Causing Fission (MeV) kcalc knormal calc HEU-SOL-THERM-046 1 89.84 Solution UO 2 SO 4 Cylinder H 2 SO 4-H 2 O / BeO & Graphite 708.24 3.6434E-03 1.01689 1.01577 0.00031 2 689.09 3.7330E-03 1.01355 1.01244 0.00032 3 677.69 3.7802E-03 1.01452 1.01341 0.00032 4 661.20 3.8213E-03 1.01508 1.01396 0.00032 5 652.90 3.9008E-03 1.01373 1.01262 0.00032 6 641.15 3.9074E-03 1.01398 1.01287 0.00034 7 622.27 4.0050E-03 1.01507 1.01395 0.00033 8 600.97 4.1664E-03 1.01414 1.01303 0.00034 9 592.65 4.2028E-03 1.01392 1.01281 0.00032 10 560.50 4.3709E-03 1.01105 1.00994 0.00036 11 530.86 4.6273E-03 1.01447 1.01336 0.00035 12 487.52 4.8643E-03 1.01375 1.01264 0.00035 13 456.65 5.1567E-03 1.01377 1.01266 0.00036
3.0 REFERENCES
- 1. International Handbook of Evaluated Criticality Safety Benchmark Experiments , NEA/NCS/DOC(95)03, Organization for Economic Cooperation and Development, September 2011 Edition (experiment HEU-SOL-THERM-046). 2. MCNP 5 and the ENDF/B
-VI Cross Section Library Homogeneous Uranium Systems Validation, NSA-TR-07-08, Rev. 0, August 20, 2007.
- 3. Software Quality Assurance Implementation for MCNP Version 5 Nuclear Criticality Safety Software for the NSA Computer System, NSA-CS-06, Rev. 1, June 2007.
4.0 ANALYSIS/PROCESS METHODOLOGY This DAC does not model a process. See Section 8.1 for description of the models and calculation methodology.
5.0 COMPUTER
CODES USED IN DAC The MCNP 5 code provides a method of analysis for criticality and shielding analysis on workstations or personal computers (PCs). MCNP 5 is one of the code s chosen for criticality safety use.
The MCNP 5 (build 1.4) code is executed on the NSA servers using the Fedor Linux operating system identified in Reference 3.
NSA-DAC-SHN-13-02 Rev. 1 Page 6 of 13 The default library is used for the critical experiment calculations.
This is primarily the Evaluated Neutron Data File B
-VI (ENDF/B
-VI) library which contains data for all nuclides (more than 300). Table 2 lists the specific elements used in this evaluation. Where the default library does not contain a "natural" mixture of isotopes, the isotopic fractions are included. The light water (lwtr.60t), beryllium oxide (beo.60t) and graphite (grph
.60t) s are used.
Table 2. Default Library Definitions for Various Elements Element ZAID* Isotopic Fraction Element ZAID Isotopic Fraction Hydrogen 1001 Nickel 28058 0.682737 Beryllium 4009 28060 0.261053 Boron 5010 0.199 28061 0.011263 5011 0.801 28062 0.035895 Carbon 6000 28064 0.009053 Nitrogen 7014 Copper 29063 0.6917 Oxygen 8016 29065 0.3083 Fluorine 9019 Zirconium 40000 Sodium 11023 Molybdenum 42000 Magnesium 12000 Silver 47107 0.5184 Aluminum 13027 47109 0.4816 Silicon 14000 Cadmium 48106 0.0 125 Sulfur 16032 48108 0.00 89 Chlorine 17000 48110 0.1249 Calcium 20000 48111 0.128 Titanium 22000 48112 0.2413 Vanadium 23000 48113 0.1222 Chromium 24050 0.043474 48114 0.2873 24052 0.837895 48116 0.0 749 24053 0.095000 Tin 50000 24054 0.023632 Samarium 149 62149 Manganese 25055 Lead 82206 0.2444 Iron 26054 0.059006 82207 0.2241 26056 0.917181 82208 0.5314 26057 0.021007 Uranium 92234 0.00 77 26058 0.002806 92235 0.8984 92236 0.00 18 92238 0.0 920
- ZAID notation for MCNP material input: Z = Atomic number; A = mass number
NSA-DAC-SHN-13-02 Rev. 1 Page 7 of 13 6.0 ASSUMPTIONS & OPEN ITEMS
6.1 Assumptions
The modeling assumptions made for the critical experiments are consistent with those described in Reference 1.
6.2 Open Items There are no open items.
7.0 ACCEPTANCE
CRITERIA
7.1 Biases
and Uncertainties The code methodology benchmark bias and bias uncertainties are not used to determine a process limit or margin of safety in this analysis and are not used in this analysis.
This analysis provides justification that the presence of uranium sulfate does not adversely affect the bias and bias uncertainties determined in Reference 2.
7.2 Area of Applicability (AoA)
This DAC extends the AoA derived in Reference 2 to include uranium sulfate solution as discussed in Section 8.0. The AoA derived in Reference 2 includes most of the materials in this study (with exception of the uranium sulfate addressed in this DAC), including uranium enrichment, elements and materials modeled, and average energy of neutrons causing fission.
Based on the previous work and the results of this DAC, it is judged that the models of this study are within the AoA of the code validation.
8.0 CALCULATIONS
8.1 Method
Discussion Materials validation is, in effect, the evaluation of how well the use of the cross section libraries esti mates t h e actual ne u tron i nteractions for t h e isotopes m aking up the m aterial. A n observable difference between calculated and measured k e f f for the bench mark experi ments would indicate a possible bias due to the system m a terials. The eval uation herein exa mines the k e f f res u lts for selected bench mark experi ments in Reference 1 (experiment HEU-SOL-THERM-046) containing uranium sulfate solution. A summary of the experi ment in f or mation is listed in Ta b l e 3. The calc ulated k e f f for these bench mark experi ments are then co mpared with si milar 235 U syste m s to esti mate the possible change in the bias.
NSA-DAC-SHN-13-02 Rev. 1 Page 8 of 13 Table 3. HEU-SOL-THERM-046 Benchmark Critical Experiment Summary Case 235 U Enrichment Uranium Form Geometry Moderator / Reflector H/235 U Other Materials kexp a expHEU-SOL-THERM-046 1 89.84 Solution UO 2 SO 4 Cylinder H 2 SO 4-H 2O / BeO & Graphite 708.24 Zircalloy, Al alloy 1.0011 0.0029 2 689.09 1.0011 0.0029 3 677.69 1.0011 0.0029 4 661.20 1.0011 0.0029 5 652.90 1.0011 0.0030 6 641.15 1.0011 0.0029 7 622.27 1.0011 0.0031 8 600.97 1.0011 0.0032 9 592.65 1.0011 0.0037 10 560.50 1.0011 0.0029 11 530.86 1.0011 0.0028 12 487.52 1.0011 0.0029 13 456.65 1.0011 0.0030 a Reference 1, Table 37.
8.2 Input
All input is obtained from Reference 1, experiment HEU-SOL-THERM-046. Beryllium Oxide and Graphite Reflected Uranyl Sulfate (HEU-SOL_THERM-0 46) Thirteen experiments involved a beryllium oxide
- and graphite-reflected tank of uranyl sulfate (UO 2 SO 4). The core was comprised of the fissile solution inside a cylindrical tank of Zircaloy
-2. The inner tank was 25 cm in diameter and 30 cm in height, with a conical bottom, surrounded by an outer aluminum
-alloy tank, and reflected by a minimum of 27.5 cm of beryllium oxide followed by a layer of at least 50 cm of graphite. The reflectors contained structural plates of aluminum alloy, as well as several penetrations throughout to accommodate the fill tube, experimental channels, control rods, safety rods, etc. The range of the uranium 235 U concentration was 36.588 to 56.51 g/liter.
8.3 Evaluations, Analysis, and Detailed Calculations The evaluation herein see k s to estimate the p o t enti a l bi a s f or syste m s of uranium sulfate syste ms. While several bench mark experi ments are available for other uranium mixtures and solutions, th e re is o n ly a limited set of uranium sulfate experi ments. Therefore, only li mited conclusions are possible. Note: A second experiment series involving uranium sulfate solution is included in Reference 1 (IEU-SOL-THERM-001) but was judged unusable due to unacceptable keff results that are not explained by the evaluators / authors of the benchmark discussion and model.
The MCNP 5 (build 1.40) code is executed on the NSA servers using the Fedor Linux operating system identified in Reference 2. Each case was run for sufficient neutron generations and neutrons per generation calc) less than 0.001. The
NSA-DAC-SHN-13-02 Rev. 1 Page 9 of 13 calculation results are recorded in Table 1.
As not all ex peri ments are exac t ly c ritical t h e calculated k e f f is nor malized to the esti mated experiment k e f f: k no rm a l = kcalc/k ex p; kexp is listed in Table 1. The data are plotted as knormal vs H/235U, knormal vs ANECF (average neutron energy causing fission) in Figures 1 and 2.
Figure 1: knormal vs H/235 U Figure 2: knormal vs Average Neutron Energy Causing Fission
The uranium sulfate solutio n s a r e nea r ly pure 2 3 5 U and are there fore most appropriately co mpared with the highly enriched 235 U (HEU) solution syste m s from Reference 2. As shown in Table 4 the UO 2 SO 4 solutions exhibit a slight positi v e bias (tendency to over 1.009001.010001.01100 1.01200 1.013001.01400 1.015001.01600 1.01700400.00450.00500.00550.00600.00650.00700.00750.00 k-normal H/U-235 1.009001.010001.01100 1.01200 1.013001.014001.01500 1.016001.017003.00E-033.50E-034.00E-034.50E-035.00E-035.50E-03 k-normal Average Neutron Energy Causing Fission (MeV)
NSA-DAC-SHN-13-02 Rev. 1 Page 10 of 13 predict k ef f), however as the a vera ge total uncertainty (total) herein is 0.00033 there is no statistically signi fica n t bias. Additi onally, a positive bi a s is conservative f or criticality sa fety. Therefore, it is concluded that no additional A OA margin is required with the use of ho mogen e ous uranium sulfate solutions
. Table 4: Average Value Comparison Group Ave knor mal MCNP Uranium Sulfate Solutions 1.0130 Ref. 2 LEU Ho mogeneous Solids 1.0046 Ref. 2 LEU Ho mogeneous Solutions 1.0001 Ref. 2 HEU Ho mogeneous Solutions 1.0003 NSA-DAC-SHN-13-02 Rev. 1 Page 11 of 13 A PPENDIX 1: REPRESENTATIVE INPUT FILES HEU-SOL-THERM-046 Proserpine case1, 36.588g U235/L (Table 1) c Benchmark HEU
-SOL-THERM-046, h(crit) = surface 1 = Table 1 & 32 c Dimensions & materials from HEU
-SOL-THERM-046 sections 3.1
- 3.3 1 1 0.100174 1 imp:n=1 $ solution in cylindrical part of tank 2 1 0.100174
-3 imp:n=1 $ solution in tapered base of tank 3 1 0.100174
-4 imp:n=1 $ solution in drain tube 4 2 4.3243E
-02 3 -5 7 imp:n=1 $ Zirca loy-2 tapered base of tank 5 2 4.3243E
-02 2 3 -6 8 14 imp:n=1 $ Zircaloy
-2 cylindrical part of tank 6 2 4.3243E
-02 4 -7 imp:n=1 $ Zircaloy
-2 drain tube 7 3 5.3248E 8 1 imp:n=1 $ air above solution 8 3 5.3248E 9 6 5 10 imp:n=1 $ air around Zircaloy
-2 tank 9 3 5.3248E 10 5 7 imp:n=1 $ air around Zircaloy
-2 tube 10 4 5.9544E 11 9 12 imp:n=1 $ Al
-alloy tank 11 4 5.9544E 12 10 imp:n=1 $ Al
-alloy tube 12 4 5.9544E 13 11 imp:n=1 $ Al
-alloy tank lip 13 2 4.3243E 14 8 imp:n=1 $ Zircaloy
-2 tank lip 14 5 7.7673E 15 8 imp:n=1 $ Teflon gasket/tank lip 15 4 5.9544E 16 17 19 21 imp:n=1 $ Al
-alloy tank cover 16 4 5.9544E 17 18 imp:n=1 $ Al
-alloy center top tube (#1) 17 4 5.9544E 19 20 imp:n=1 $ Al
-alloy center top tube (#2) 18 4 5.9544E 21 22 imp:n=1 $ Al
-alloy center top tube (#3) 19 3 5.3248E 18 imp:n=1 $ air in tube #1 20 3 5.3248E 20 imp:n=1 $ air in tube #2 21 3 5.3248E 22 imp:n=1 $ air in tube #3 30 6 1.4204E 30 11 12 13 14 15 16 17 19 21
41 43 47 49 50 51 imp:n=1 $ BeO section #1 31 6 1.4204E 31 30 12 17 19 21
41 43 47 49 imp:n=1 $ BeO section #2 32 6 1.4204E 32 31 41 43 imp:n=1 $ BeO section #3 33 6 1.4204E 33 17 19 21 imp:n=1 $ BeO top section #4 34 4 5.9544E 34 17 19 21 imp:n=1 $ Al
-alloy BeO cover 35 7 8.2737E 35 12 17 19 21 30 31 32 33 34 41 43 44 45 47 49 50 51 52 53 imp:n=1 $ Graphite 40 3 5.3248E 40 imp:n=1 $ air, reflector channel #1 41 4 5.9544E 41 40 imp:n=1 $ Al
-alloy, reflector channel #1 42 3 5.3248E 42 imp:n=1 $ air, reflector channel #2 43 4 5.9544E 43 42 imp:n=1 $ Al
-alloy, reflector channel #2 44 3 5.3248E 44 imp:n=1 $ air, reflector channel #3 45 3 5.3248E 45 imp:n=1
$ air, reflector channel #4 46 3 5.3248E 46 imp:n=1 $ air, reflector channel #5 47 4 5.9544E 47 46 imp:n=1 $ Al
-alloy, reflector channel #5 48 3 5.3248E 48 imp:n=1 $ air, reflector channel #6 49 4 5.954 4E-02 -49 48 imp:n=1 $ Al
-alloy, reflector channel #6 50 3 5.3248E 50 11 imp:n=1 $ air, reflector channel #7 51 3 5.3248E 51 11 imp:n=1 $ air, reflector channel #8 52 3 5.3248E 52 imp:n=1 $ air, ref lector channel #9 53 3 5.3248E 53 imp:n=1 $ air, reflector channel #10 99 0 35 imp:n=0 $ outside world
1 pz 23.875 $ critical h of solution from base of tapered base 2 rcc 0 0 0.87 0 0 23.005 12.5 $ solution in cylindrical part of tank (max critical h) 3 trc 0 0 0 0 0 0.87 0.3 12.5 $ solution in tapered base 4 rcc 0 0 0 0 0
-80.5 0.3 $ solution in drain tube 5 trc 0 0
-0.0931 0 0 0.8 7 0.4 12.6 $ Zircaloy
-2 tank tapered base 6 rcc 0 0 0.7769 0 0 29.0 12.6 $ Zircaloy
-2 cylindrical tank 7 rcc 0 0 0 0 0
-80.5 0.4 $ Zircaloy
-2 drain tube 8 rcc 0 0 30.7 0 0
-15.16 12.5 $ air above solution (min criti cal h) 9 rcc 0 0
-0.3 0 0 30.0 12.9 $ air around Zircaloy
-2 tank 10 rcc 0 0 0 0 0
-80.5 0.7 $ air around Zircaloy
-2 drain tube 11 rcc 0 0
-0.5 0 0 30.2 13.1 $ Al
-alloy tank 12 rcc 0 0
-0.3 0 0
-80.2 0.9 $
Al-alloy drain tube 13 rpp -15 15 -15 15 29.5 29.7 $ Al
-alloy tank lip 14 rpp -15 15 -15 15 29.7 29.8 $ Zircaloy
-2 tank lip 15 rpp -15 15 -15 15 29.8 30.7 $ Teflon gasket/tank lip
NSA-DAC-SHN-13-02 Rev. 1 Page 12 of 13 16 rpp -15 15 -15 15 30.7 33.5
$ Al-alloy tank cover 17 rcc 0 0 30.7 0 0 80 1.0 $ Al
-alloy center top tube #1 OD 18 rcc 0 0 30.7 0 0 80 0.9 $ Al
-alloy center top tube #1 ID 19 rcc 10 0 30.7 0 0 80 1.0 $ Al
-alloy 2nd 2cmOD top tube #2 20 rcc 10 0 30.7 0 0 80 0.9 $ Al
-alloy 2nd 2cmOD top tube #2 ID 21 rcc -10 0 30.7 0 0 80 0.5 $ Al
-alloy 1cmOD top tube #3 22 rcc -10 0 30.7 0 0 80 0.4 $ Al
-alloy 1cmOD top tube #3 ID 30 rpp -40 40 -25 25 -20.5 50.7
$ BeO reflector section #1 31 rpp -30 30 -35 35 -30.5 60.7 $ BeO reflector section #2 32 rpp -20 20 -40 40 -30.5 60.7 $ BeO reflector section #3 33 rpp -15 15 -15 15 60.7 63.2 $ BeO reflector section #4 (top/cover) 34 rpp -20 20 -20 20 63.2 65.2 $ Al
-alloy BeO reflector top/cover 35 rpp -90 90 -90 90 -80.5 110.7 $ Graphite reflector 40 rcc 15
-90 4.5 0 180 0 1.5 $ air, y
-axis channel #1 ID 41 rcc 15 -90 4.5 0 180 0 1.7 $ Al
-alloy tube, y
-axis channel #1 OD 42 rcc 90 4.5 0 180 0 1.5 $ air, y
-axis channel #2 ID 43 rcc 90 4.5 0 180 0 1.7 $ Al
-alloy tube, y
-axis channel #2 OD 44 rcc 50
-90 4.5 0 180 0 2.5 $ air, y
-axis channel #3 ID 45 rcc 90 4.5 0 180 0 2.5 $ air, y
-axis channel #4 ID 46 rcc 25
-90 14.5 0 180 0 0.6 $ air, y
-axis channel #5 ID 47 rcc 25
-90 14.5 0 180 0 0.8 $ Al
-alloy tube, y
-axis channel #5 OD 48 rcc 90 14.5 0 180 0 0.6 $ air, y
-axis channel #6 ID 49 rcc 90 14.5 0 180 0 0.8 $ Al
-alloy tube, y
-axis channel #6 OD 50 rcc -90 0 9.5 180 0 0 0.9 $ air, x
-axis channel #7 ID 51 rcc -90 0 19.5 180 0 0 0.6 $ air, x
-axis channel #8 ID 52 rcc -90 50 9.5 180 0 0 2.5 $ air, x
-axis channel #9 ID (assumed z, dimension not shown) 53 rcc 50 9.5 180 0 0 2.5 $ air, x
-axis channel #10 ID (assumed z, dimension not shown) c UO2SO4 solu tion (mass density = 1.05607 g/cc) m1 92234.60c 8.0385E
-07 92235.60c 9.3399E
-05 92236.60c 1.8632E
-07 92238.60c 9.4426E
-06 1001.60c 6.6149E
-02 8016.60c 3.3787E
-02 16032.60c 1.3383E
-04 mt1 lwtr.60t c c Zircaloy
-2 (mass density = 6.56 g/cc) m2 40000.60c 4.2552E
-02 $ zirconium 50000 4.8254E
-04 $ tin 26054.60c 5.8429E
-06 $ iron 26056.60c 9.0833E
-05 $ iron 26057.60c 2.0797E
-06 $ iron 26058.60c 2.7729E
-07 $ iron 24050.60c 3.3012E
-06 $ chromium 24052.60c 6.3661E
-05 $ chromium 24053.60c 7.2178E
-06 $ chromium 24054.60c 1.7969E
-06 $ chromium 28058.60c 2.2977E
-05 $ nickel 28060.60c 8.78 42E-06 $ nickel 28061.60c 3.8031E
-07 $ nickel 28062.60c 1.2083E
-06 $ nickel 28064.60c 3.0627E
-07 $ nickel c c air m3 7014.60c 4.1985E
-5 $ nitrogen 8016.60c 1.1263E
-5 $ oxygen c c AL-alloy (mass density = 2.66 g/cc) m4 13027.60c 5.7529E
-02 $ aluminum 12000.60c 1.9772E
-03 $ magnesium 26054.60c 6.0080E
-07 $ iron 26056.60c 9.3398E
-06 $ iron 26057.60c 2.1384E
-07 $ iron 26058.60c 2.8512E
-08 $ iron 14000.60c 1.9962E
-05 $ silicon 22000.60c 2.0074E
-06 $ titanium
NSA-DAC-SHN-13-02 Rev. 1 Page 13 of 13 25055.60c 8.7474E
-07 $ manganese 23000.60c 9.4337E
-07 $ vanadium 29000 4.2870E
-06 $ Cu substituted for Zn c c Teflon (mass density = 2.15 g/cc) m5 6000.60c 2.5891E
-02 $ carbon 9019.60c 5.1782E
-02 $ fluorine c c BeO (mass density = 2.95 g/cc) m6 4009.60c 7.0937E
-02 $ beryllium 8016.60c 7.0937E
-02 $ oxygen 13027.60c 9.8763E
-06 $ aluminum 26054.60c 9.3840E
-08 $ iron 26056.60c 1.4588E-06 $ iron 26057.60c 3.3401E
-08 $ iron 26058.60c 4.4534E
-09 $ iron 14000.60c 3.1627E
-06 $ silicon 12000.60c 3.6546E
-07 $ magnesium 25055.60c 3.2337E
-08 $ manganese 48106.66c 5.9265E
-11 $ cadmium 48108.66c 4.2197E
-11 $ cadmium 48110.66c 5.9218E
-10 $ cadmium 48111.66c 6.0687E
-10 $ cadmium 48112.66c 1.1441E
-09 $ cadmium 48113.66c 5.7937E
-10 $ cadmium 48114.66c 1.3621E
-09 $ cadmium 48116.66c 3.5512E
-10 $ cadmium 47107.60c 4.2281E
-09 $ silver 47109.60c 3.9282E
-09 $ silver 29063.60c 1.9337E
-07 $ copper 29065.60c 8.6188E
-08 $ copper 82206.60c 6.2865E
-09 $ lead 82207.60c 5.7643E
-09 $ lead 82208.60c 1.3669E
-08 $ lead 28058.60c 6.1995E
-08 $ nickel 28060.60c 2.3701E
-08 $ nickel 28061.60c 1.0261E
-09 $ nickel 28062.60c 3.2600E
-09 $ nickel 28064.60c 8.2636E
-10 $ nickel 6000.60c 1.4791E
-04 $ carbon 5010.60c 9.8103E
-09 $ boron 5011.60c 3.9488E
-08 $ boron mt6 beo.60t c c Graphite (mass density = 1.65 g/cc)
m7 6000.60c 8.2718E
-02 $ carbon 5010.60c 3.6580E
-09 $ boron 5011.60c 1.4724E
-08 $ boron 10 01.60c 1.6267E
-05 $ hydrogen 11023.60c 8.2120E
-07 $ sodium 20000.60c 7.4378E
-07 $ calcium 62149.66c 1.1343E
-10 $ samarium 149 22000.60c 1.0376E
-07 $ titanium 26054.60c 2.0995E
-09 $ iron 26056.60c 3.2639E
-08 $ iron 26057.60c 7.4729E
-10 $ iron 26058.60c 9.9638E
-11 $ iron 42000.60c 1.0470E
-07 $ molybdenum 23000.60c 6.2418E
-07 $ vanadium 17000.60c 9.8095E
-08 $ chlorine mt7 grph.60t c kcode 10000 1.0 15 500
ksrc 0 0 5