ML22340A008

Appendix B, Activation Study of Components of the Aerotest Radiography and Research Reactor in Support of Its Characterization - Redacted

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Issue date:	09/19/2011
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Activation Study of Components of the Aerotest Radiography and Research Reactor in Support Of Its Characterization September 19, 2011 Prepared by Nolan E. Hertel, Ph.D., P.E. (Georgia) 3749 Allenhurst Drive Norcross, GA 30092 For EnergySolutions

Rev. 0.0 Table of Contents Calculation Package Cover Sheet ....................................................................................................1 Design Review Checklist .................................................................................................................2 Table of Contents .............................................................................................................................3 List of Tables ...................................................................................................................................4

1. Introduction ..............................................................................................................................6

2. Purpose .....................................................................................................................................8

3. Method of Analysis ..................................................................................................................8 3.1 Calculation of Fluxes .......................................................................................................8 3.1.1 ORIGEN-S Calculations ............................................................................................. 8 3.1.2 Use of ORIGEN-S Activation Results........................................................................ 9

4. Assumptions and Design Input...............................................................................................10 4.1 Design Input ...................................................................................................................10 4.1.1 Neutron Fluxes .......................................................................................................... 10 4.1.2 ARRR Operating History.......................................................................................... 10 4.1.3 Composition and Isotopic Data for ORIGEN-S Analysis ........................................ 10 4.2 Design Assumptions ......................................................................................................20 4.2.1 Operating Assumptions ............................................................................................. 20 4.2.2 Modeling Assumptions ............................................................................................. 20

5. Neutron Flux Calculations......................................................................................................22 5.1 ARRR Core ....................................................................................................................22 5.2 ARRR Criticality ...........................................................................................................22 5.3 Experimental Facilities ..................................................................................................23 5.4 Core Shroud/Structure ...................................................................................................23 5.5 Reactor pool and Surrounding Structures ......................................................................26

6. Activation Product Calculations .............................................................................................26 6.1 ORIGEN-S Activation and Decay Calculations ............................................................26 6.1.1 ORIGEN-S ................................................................................................................ 26 6.1.2 Irradiation Scenario ................................................................................................... 27

7. Activation Results ..................................................................................................................27 7.1 Core Support Structure. .................................................................................................27 7.1.1 Aluminum Activities................................................................................................. 27 7.1.2 Bolts in T-Bars Activities ......................................................................................... 34 7.2 Instrument Guide Tubes .................................................................................................34 7.3 Thermal Column and Neutron Radiography Assembly.................................................34 7.3.1 Graphite Activities .................................................................................................... 34 7.3.2 Aluminum Activities................................................................................................. 34 7.3.3 Lead Activities .......................................................................................................... 39 7.3.4 Boral Shutter Activities............................................................................................. 39 7.4 Aluminum Tank Activities ............................................................................................39 7.5 Bioshield Activities ........................................................................................................39 7.5.1 Concrete Activities.................................................................................................... 39 7.5.2 Rebar Activities ........................................................................................................ 39 7.6 Soil Activities.................................................................................................................39 Rev. 0 3/11/03 2 of 47

Rev. 0.0

8. Summary ................................................................................................................................43

9. References ..............................................................................................................................44

10. Spreadsheets ...........................................................................................................................46 Rev. 0 3/11/03 3 of 47

Rev. 0.0 List of Tables Table 1. MCNP Energy Groups ....................................................................................................10 Table 2. Yearly Operating History 1965-2011 .............................................................................11 Table 3. Composition of Regulatory Concrete [4].......................................................................12 Table 4. Range of Impurities from NUREG/CR-3474 for ...........................................................12 Table 5. High and Average Composition for ................................................................................13 Table 6. Lead Composition [8] .....................................................................................................14 Table 7. Aluminum Composition for Activation Calculations [15] .............................................14 Table 8. SCALE SS304 Composition Used in Flux Calculations [4] ..........................................16 Table 9. SS304 Composition with Impurities Based on Table 8 and NUREG/CR-3474 [7] ........16 Table 10. Soil Composition Used in Flux Calculations.[11] ........................................................17 Table 11. Air Composition [16] .....................................................................................................17 Table 12. Graphite Composition [10] ............................................................................................17 Table 13. Aluminum Fuel Compositions. ......................................................................................18 Table 14. Stainless Steel Fuel Compositions ................................................................................18 Table 15. Rebar Constituents for Activation Calculation [7] ........................................................21 Table 16. Boral Composition for Activation Calculations. ..........................................................21 Table 17. Irradiation Scenario Based on Reactor Operating History ...........................................28 Table 18. Description of Locations in Figure 9 and Figure 10. ....................................................31 Table 19. Activation Levels as of October 1, 2012 in the Core Shroud, Top Grid Plate, ............32 Table 20. T-Beam Activities (Ci/g) on October 1, 2012. See Figure 7 for the locations of the cells..........................................................................................................................................33 Table 21. Activities (Ci/g) in Representative Bolts in the T-Beams of the Structure on October 1, 2012. The bolts were assumed to be SS304. ..........................................................................35 Table 22. Activities (Ci/g) in the Instrument Guide Tubes on October 1, 2012...........................36 Table 23. Activities in the Graphite (Ci/g) in the Thermal Column and Neutron Radiography Assembly on October 1, 2012. ................................................................................................37 Table 24. Activities (Ci/g) in the Aluminum in Representative Locations of the Thermal Column and the Neutron Radiography Assembly on October 1, 2012.................................................38 Table 25. Activities in the Lead (Ci/g) at the Front and Back ......................................................40 Table 26. Boral Shutter Activities (Ci/g) on October 1, 2012 ......................................................41 Table 27. Aluminum Tank Activities (Ci/g) on October 1, 2012. ................................................41 Table 28. Activities in the Concrete Below and to the Nearest Side of the Core (Ci/g) on October 1, 2012. ......................................................................................................................42 Table 29. Activities in the Rebar Below and to the Nearest Side of the Core (Ci/g) on October 1, 2012. ........................................................................................................................................42 Table 30. Activity in Soil (Ci/g) on October 1, 2012. ..................................................................43 Rev. 0 3/11/03 4 of 47

Rev. 0.0 List of Figures Figure 1. ARRR Reactor Cut-Away. ..............................................................................................7 Figure 2. ARRR Cut-Away showing the Thermal Column and the Radiography Port. .................7 Figure 3. ARRR Aluminum Clad Fuel Element. [13] ..................................................................19 Figure 4. ARRR Stainless Steel Clad Fuel Element. [13] ............................................................19 Figure 5. June 10, 2011 Core Map Indicating Location of Various Elements. [14] .....................24 Figure 6. June 10, 2011 Core Map Indicating Location of Stainless Steel Elements. [14] ..........25 Figure 7. Vertical Cut of the ARRR MCN5 Model Indicating Where Volume-Averaged Fluxes Were Computed.......................................................................................................................29 Figure 8. Horizontal Cut of the ARRR MCN5 Model Indicating Where Volume-Averaged Fluxes Were Computed. ..........................................................................................................29 Figure 9. Horizontal Cut Showing Locations at Which Localized Fluxes Were Computed. .......30 Figure 10. Vertical Cut Showing Locations at Which Localized Fluxes Were Computed. .........30 Figure 11. ARRR Core and Support Structure. ............................................................................31 Rev. 0 3/11/03 5 of 47

Rev. 0.0

1. Introduction In this report, engineering estimates using three-dimensional Monte Carlo code simulations of fluxes in the Aerotest Radiography and Research Reactor (ARRR) were performed. These fluxes were used with activation codes to determine the activation product levels remaining in the ARRR on October 1, 2012 assuming no significant operation after September 2, 2011.

The ARRR is a 250 kW TRIGA-type reactor located in California. The reactor went critical in 1965 and has operated through June 2010. The original core was comprised of aluminum clad fuel elements and then later stainless steel clad fuel elements were added. The active dimensions of the core are 19.44 inches in diameter and 14 inches in height for the aluminum clad fuel elements and 15 inches high for the steel clad fuel elements. The shielding of the core is provided by demineralized water, concrete, lead and wood. Two cutaways of the reactor are shown in Figure 1 and Figure 2.

The MCNP code was used to compute fluxes in the various components of the system. The fluxes were binned into a fast flux group (1-10 MeV), a resonance flux group (0.625- 1 MeV),

and several thermal flux groups. The fast, resonance and total thermal flux levels in various components of the reactor were computed for use in the activation code. The fluxes were computed were for a maximum power of 250 kW. The ARRR yearly operating history was used to develop a time-dependent irradiation scenario from 1965-2011. The operational history encompasses a total burnup of 14,196,548 kW-hr over that time period.

The activation calculations were performed with the SCALE version of the well-known isotope generation and depletion code ORIGEN [1]. This code uses a thermal flux in combination with thermal, resonance and fast spectral parameters to perform the activation calculations. The reactor was assumed to operate at the maximum power level for an appropriate time period at the end of each year to account for that years total kW-hr of operation. The code was used to compute the isotope buildup and decay over the operating life of the reactor and then allowed to decay the isotopes for several decay times after shutdown. The activation product concentration levels reported herein are for decay to October 1, 2012 and assumed no significant operation after September 2, 2011. There will be minor amounts of operation to perform activities required to maintain the reactor operating license, but to provide meaningful results requires that a snapshot in time be used to compute the activation levels. In addition these small periods of operation would not add significant activity to the longer lived radionuclides of importance in decommissioning. Since no chemical analyses of the components used in the construction are available, the results reported in this work can only be considered to be estimates made using representative impurity levels in the components of the reactor. Two representative impurity levels were used for most of the components, one representing an average impurity level and the other representing an impurity level higher in the elements which drive the activation analysis.

Page 6 of 47

Rev. 0.0 Figure 1. ARRR Reactor Cut-Away.

Radia~icn Beam C'ilt-cber Shield I.

I .

' I I

. \

Figure 2. ARRR Cut-Away showing the Thermal Column and the Radiography Port.

Page 7 of 47

Rev. 0.0

2. Purpose The activation calculations are principally aimed at two objectives:

• To provide reasonable estimates of the radionuclides that are still present in sufficiently high levels to drive dose rates to personnel during dismantling of the facility, and

• To provide reasonable, yet conservative, estimates of the radionuclide inventory for use in the client determining their waste disposal need and cost.

3. Method of Analysis 3.1 Calculation of Fluxes In order to perform activation calculations, reasonable estimates of the neutron fluxes during the operation of the Aerotest Radiography and Research Reactor (ARRR). The neutron fluxes for the ARRR were computed using a three-dimensional MCNP5 [6] model of the ARRR core, major support and experimental structures where details existed. The MCNP5 code is a Monte Carlo code was used without biasing techniques for calculate volume average neutron fluxes for structures near the core.

For structures further away from the core, the MCNP weight windows generator and resulting weight windows were used to calculate the neutron fluxes. The MCNP5 variance reduction devices require an iterative approach to the solution of the transport problem and are adjusted until the best set of variance reduction parameters are obtained. Such variance reduction techniques were employed in the present work, the details of which will not be presented in this document. The mesh tally algorithm within MCNP5 was used to compute the neutron fluxes on a user-defined grid that is superimposed over the 3D model. In this work, neutron fluxes at several mesh tally locations were used throughout various structures of the ARRR. Additionally these mesh tallies were used to determine maximum neutron fluxes in smaller volumes in the structure near the core.

3.1.1 ORIGEN-S Calculations The computed fluxes based on an operating power level of 250 kW(th) were used to create the flux input data for the ORIGEN-S code [1]. The ORIGEN-S code was used to compute the activation product levels in components of interest during the 45 years of operation and on October 1, 2012. ORIGEN-S is a well-respected isotope buildup and depletion code that is widely used in the nuclear industry. The fluxes from the transport computations were used to determine the three spectral factors that are needed to properly weight the thermal, resonance and fast activation cross sections. The MCNP5-computed total thermal neutron flux is also supplied to the code as part of the ORIGEN-S neutron input data. The spectral factors were calculated using the formulas found in Reference [1]. They are reviewed in the following three subsections.

3.1.1.1 Thermal Spectral Factor The thermal spectral factor, THERM, was calculated using the following equation:

n i

0.15906 i 1 J Ei THERM (3.1) th Page 8 of 47

Rev. 0.0 where 0 625eV th = the thermal flux (= E dE ) which is obtained by summing the flux over the 0

lower energy groups of the one-dimensional calculation; i = flux in the ith group; n = the number of thermal groups; and Ei = the energy derived by some logical method for representing the energy of each group

[1]. To perform this summation with the flux data from MCNP5 computations, the logarithmic average energy for each group was used as Ei.

3.1.1.2 Resonance Region Spectral Factor The resonance region spectral factor, RES, was calculated by the following equation:

mres j RES (3.2) j 1 th where mres = number of groups over the energy range from 0.625 eV to 1 MeV.

3.1.1.3 Fast Spectral Factor The fast spectral factor, FAST, is calculated from the following equation:

kfast l

FAST l (3.3) th where kfast = number of groups above 1 MeV.

3.1.2 Use of ORIGEN-S Activation Results The activation calculations were performed for selected components of the ARRR. The parameter of interest generated by ORIGEN-S is the concentration of activation products in curies per a user-supplied basis. For this work, the number of gram-atoms each element of the component (on a per gram basis) are input with the fluxes and the reactor operating scenario to compute the activation product concentrations in Ci per gram. Attempts must be made to include trace impurities that may be of importance in activation of the component. Since code will compute activation levels that are below practical concern, a cutoff activity level (Ci/g) is often employed.

Page 9 of 47

Rev. 0.0

4. Assumptions and Design Input 4.1 Design Input 4.1.1 Neutron Fluxes One set of energy groups was used in MCNP5 to determine the computed fluxes. This energy group structure used in the MCNP5 calculations were necessary to computer determine the THERM, RES and FAST constants for input into the ORIGEN-S calculation. For the thermal

(<0.625 eV) structure multiple groups were used, one resonance group from 0.625 eV to 1 MeV and a fast group above 1 MeV.

Table 1. MCNP Energy Groups Upper Energy Group (MeV) 1 1.00E+02 2 1.00E+00 3 6.25E-08 4 1.82E-08 5 1.28E-08 6 5.76E-09 7 2.60E-09 8 1.18E-09 9 5.31E-10 10 2.40E-10 11 1.08E-10 12 4.90E-11 13 2.21E-11 14 1.00E-11 4.1.2 ARRR Operating History The ARRR operating history was taken from a spreadsheet provided by EnergySolutions from 1962 through its shutdown in 2002 is shown in Table 2.

4.1.3 Composition and Isotopic Data for ORIGEN-S Analysis In order to facilitate the transport calculation of the fluxes, different compositions were often used in the transport model than in the activation calculations. In these cases, the compositions used in the flux calculations included the principal components while the compositions used in the activation calculations were designed to provide activation levels for trace elements that might be present.

Atomic numbers (z) and atomic masses (A) were taken from Reference [5] to assist in providing the gram-atoms input for ORIGEN-S from the activation compositions.

Page 10 of 47

Rev. 0.0 Table 2. Yearly Operating History 1965-2011 Period kW Total Period kW Total Ending hours kW hours Ending hours kW hours Jun-66 38,152 38,152 Jun-89 653,637 6,524,692 Jun-67 57,949 96,101 Jun-90 724,951 7,249,643 Jun-68 80,409 176,510 Jun-91 690,150 7,939,793 Jun-69 61,744 238,254 Jun-92 676,144 8,615,937 Jun-70 140,279 378,533 Jun-93 371,146 8,987,083 Jun-71 215,123 593,656 Jun-94 344,002 9,331,085 Jun-72 208,086 801,742 Jun-95 350,692 9,681,777 Jun-73 196,161 997,903 Jun-96 352,727 10,034,504 Jun-74 165,159 1,163,062 Jun-97 340,389 10,374,893 Jun-75 203,827 1,366,889 Jun-98 447,745 10,822,638 Jun-76 201,868 1,568,757 Jun-99 369,161 11,191,799 Jun-77 244,329 1,813,086 Jun-00 275,791 11,467,590 Jun-78 266,586 2,079,672 Jun-01 240,255 11,707,845 Jun-79 293,314 2,372,986 Jun-02 211,950 11,919,795 Jun-80 289,969 2,662,955 Jun-03 222,892 12,142,687 Jun-81 281,811 2,944,766 Jun-04 315,652 12,458,339 Jun-82 275,699 3,220,465 Jun-05 284,958 12,743,297 Jun-83 309,172 3,529,637 Jun-06 216,433 12,959,730 Jun-84 374,479 3,904,116 Jun-07 252,588 13,212,318 Jun-85 445,421 4,349,537 Jun-08 326,056 13,538,374 Jun-86 484,242 4,833,779 Jun-09 279,181 13,817,555 Jun-87 487,412 5,321,191 Jun-10 290,319 14,107,874 Jun-88 549,864 5,871,055 15-Oct-10 88154 14,196,028 02-Sep-11 520 14,196,548 4.1.3.1 Concrete For the flux computations, the composition of regulatory concrete as given in the SCALE standard composition manual [4] was used. That composition is given in Table 3. The density was 2.3 g/cm3.

Representative impurity levels for power reactor bioshields (none having iron-ore aggregate) from NUREG/CR-3474 [7] are shown in Table 4. More impurities are listed in that NUREG reference, but only those potential drivers of the activation products of concern are in the table. The impurities in concrete were coupled with the principal components of NIST concrete and the resulting weight fractions renormalized to unity to form the average and high impurity bioshield concrete concentrations to be used in the activation calculations. Those compositions are shown in Table 5.

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Rev. 0.0 Table 3. Composition of Regulatory Concrete [4]

Element Atomic Number Weight Fraction O 8 0.532 Si 14 0.337 Ca 20 0.044 Al 13 0.034 Na 11 0.029 Fe 26 0.014 H 1 0.01 Table 4. Range of Impurities from NUREG/CR-3474 for Examined Bioshield Concrete Compositions [7]

Element Average ppm High ppm Ti 2121 7900 Cr 109 540 V 103 490 Mn 377 990 P 5000 5000 N 120 120 Cl 45 59 Sr 438 940 Ba 950 7060 Eu 0.98 3.1 Co 9.8 31 Ni 38 87 Cu 25 60 Zn 75 340 Page 12 of 47

Rev. 0.0 Table 5. High and Average Composition for Bioshield Concrete Activation Calculations.

Average Weight High Impurity Element Fraction Weight Fraction H 9.91E-03 9.77E-03 O 5.27E-01 5.20E-01 Na 2.87E-02 2.83E-02 Al 3.37E-02 3.32E-02 Si 3.34E-01 3.29E-01 Ca 4.36E-02 4.30E-02 Fe 1.39E-02 1.37E-02 Ti 2.10E-03 7.72E-03 Cr 1.08E-04 5.28E-04 V 1.02E-04 4.79E-04 Mn 3.73E-04 9.67E-04 P 4.95E-03 4.88E-03 N 1.19E-04 1.17E-04 Cl 4.46E-05 5.76E-05 Sr 4.34E-04 9.18E-04 Ba 9.41E-04 6.90E-03 Eu 9.71E-07 3.03E-06 Co 9.71E-06 3.03E-05 Ni 3.76E-05 8.50E-05 Cu 2.48E-05 5.86E-05 Zn 7.43E-05 3.32E-04 Total 1.00E+00 1.00E+00 4.1.3.2 Lead

• For transport computations, the material was treated as pure Pb with a density of 11.4 gm/cm3.

• The composition of lead (density of 11.4 g/cm3) used in the activation calculation is shown in Table 6. It is taken from the ASM Metals Reference Book composition for common lead [8]. The 0.2% maximum weight fraction reported there for the sum of Sn, Sb, and As was equally divided up among those elements to create the table. Since a range of impurities was not available, only one composition was used in the activation calculation.

4.1.3.3 Aluminum

• For the flux calculations, the composition of aluminum metal was taken as pure aluminum. The density of aluminum is 2.7 g/cm3 [4].

• The composition with impurities is shown in Table 7. This compositions were based on Reference [15].

Page 13 of 47

Rev. 0.0 Table 6. Lead Composition [8]

Weight Element Fraction Zn 1.00E-05 Fe 2.00E-05 Ag 5.00E-05 Cu 1.50E-05 As 6.67E-06 Sb 6.67E-06 Sn 6.67E-06 Bi 5.00E-04 Pb 9.99E-01 Table 7. Aluminum Composition for Activation Calculations [15]

Element High PPM Average PPM Be 10 5.5 Ba 10 5.05 C 100 50.05 Ca 50 25.05 Cd 5 2.505 Cl 10 5.05 Co 5 2.55 Cr 50 26 Cu 100 52.5 Fe 500 450 Ga 200 105 Li 10 5.5 Mg 50 27.5 Mn 50 27.5 N 7 4 Na 500 250.05 Ni 20 10.5 O 100 50.5 P 30 15.5 Pb 50 25.5 S 20 10.1 Si 1000 600 Sn 30 15.05 Ti 100 55 V 100 52.5 Zn 200 105 Zr 40 25 Al 996653 997991.545 Page 14 of 47

Rev. 0.0 4.1.3.4 Stainless Steel 304

• The composition used for stainless steel SS304 in the flux computations was taken from the SCALE standard composition library [4]. This composition was used for steel in the steel clad fuel for the calculations. It is shown in Table 8.

• Impurity data were taken from NUREG/CR-3474 and used to create the SS304 compositions for activation calculations. Those compositions are shown in Table 9.

Since the fuel will be removed, no activation calculation was performed for the fuel. The SS304 composition will be used to compute some approximate activities for steel bolts that may be near the core.

o The average is based on the average impurity composition of SS304 samples analyzed in the NUREG. The high impurity values are the highest value reported in that document.

o For the high impurity composition, the weight fractions were renormalized so they summed to 1.0. Hence, the reduced fractions of Fe, Cr, Mn, Ni, Ti and Eu in the high-impurity concentration when compared to the average composition.

• The density used for SS304 in these calculations was 7.94 g/cm3 from Ref. [4].

4.1.3.5 Soil The generic soil composition was based on Reference [11] for the flux computations, see Table 10. The composition used from that reference represents an average of U.S. soils.

Any elements with a weight percent of less than 0.1% were not included in the flux calculations. A density of 1.6 g.cm3 was used. The soil was assumed to be dry.

Since no composition was available for red clay, the average soil composition was used in the activation calculations as well.

4.1.3.6 Air Air was only used in the flux computation and was taken from the NIST ASTAR and PSTAR composition library. [16] The density used was 1.23(10-2) g/cm3.

4.1.3.7 Graphite The flux calculations were performed with graphite as pure carbon with a density of 2.3 g/cm3.

The graphite composition used in the activation calculations is shown in and was taken from Reference [10]. The values are representative of thermally purified or AGOT graphite. If graphite is thermally purified, rare earths like europium can remain behind in greater quantities than if chemically purified. Since no range of values for europium or chlorine was present in the reference, 1ppm was added to the composition. The activation results can be scaled from that level.

4.1.3.8 Boron Carbide The B4C in the control rods was assumed to be pure and only used in the flux calculations. Its density was 2.52 g/cm3.

Page 15 of 47

Rev. 0.0 Table 8. SCALE SS304 Composition Used in Flux Calculations [4]

Weight Element Fraction Fe 0.68375 Cr 0.19 Ni 9.5E-02 Mn 2.0E-02 Si 1.0E-02 C 8.0E-04 P 4.5E-04 Table 9. SS304 Composition with Impurities Based on Table 8 and NUREG/CR-3474 [7]

"Average" High-Impurity Element Weight Weight Fraction Fraction Fe 6.96E-01 6.86E-01 Cr 1.81E-01 1.79E-01 Ni 9.85E-02 9.72E-02 Mn 1.48E-02 1.46E-02 Cu 3.03E-03 8.15E-03 Mo 2.56E-03 5.50E-03 Co 1.39E-03 2.57E-03 Ti 5.91E-04 6.00E-04 Zn 4.50E-04 2.23E-03 V 4.49E-04 6.90E-04 N 4.45E-04 5.25E-04 Ce 3.66E-04 5.50E-04 As 1.91E-04 1.01E-03 W 1.83E-04 5.20E-04 Ga 1.27E-04 4.50E-04 Eu 1.97E-08 2.00E-08 Nb 8.77E-05 3.00E-04 Page 16 of 47

Rev. 0.0 Table 10. Soil Composition Used in Flux Calculations.[11]

Element Weight Fraction O 0.502 Si 0.265 Al 0.067 Fe 0.055 Ti 0.0045 Ca 0.05 Mg 0.013 K 0.014 Na 0.006

]

Table 11. Air Composition [16]

Weight Element Fraction C 0.00012 N 0.755 O 0.232 Ar 0.0128 Table 12. Graphite Composition [10]

Weight Element Fraction C 1.000E+00 Ca 1.470E-04 Si 4.600E-05 V 2.500E-05 S 1.900E-05 Ti 1.100E-05 Fe 1.000E-05 Al 2.500E-06 B 4.000E-07 Eu 1.000E-06 Cl 1.000E-06 Page 17 of 47

Rev. 0.0 Table 13. Aluminum Fuel Compositions.

Element Weight Fraction H 0.0099728 Zr 0.9100272 U-235 0.01 U-238 0.07 Table 14. Stainless Steel Fuel Compositions Element Weight Fraction H 0.0543098 Zr 0.08256902 U-235 0.015 U-238 0.105 4.1.3.9 Water The water in the reactor tank was assumed to be pure H2O at a density of 1.0 g/cm3. It was only used in the flux computations.

4.1.3.10 Fuel There were two types of fuel elements used in the reactor during its operation: aluminum clad and stainless steel clad. Diagrams of these fuel elements are shown in Figure 3 and Figure 4.[13]

4.1.3.10.1 Fuel meat The meat of the aluminum fuel elements was assumed to have a density of 5.61 g/cm3 and their composition is given in Table 13.

The meat of the stainless steel fuel elements was assumed to have a density of 5.61 g/cm3 and their composition is given in Table 14.

No activation product or fission product buildup calculations were done for the fuel.

4.1.3.10.2 Cladding The fuel elements were clad in SS304 or aluminum, as appropriate, using the previous compositions of those materials.

4.1.3.10.3 Samarium Oxide The samarium oxide burnable poison was assumed to be pure and only used in the flux calculations. Its density was 7.93 g/cm3.[17]

Page 18 of 47

Rev. 0.0 ALU~UNUM 'l'OP

~/ ENO-f'ltX'l'UR.E GRAPHITE J;£

• _L_S?.\CER d RURNAl:IJ.,E POISON Z8, J 7 IN.

ZI RCONIUM HYDRIDE~

8 WT URANIUM or, l 11 lN.

l. 47 IN,

l. -ti IN.

0 l

J CRAPl"JTE""'I - T

4. 0 JN.

_J._

I BOTTOM £NO-FIXTURE Figure 3. ARRR Aluminum Clad Fuel Element. [13]

STAIIILCSS SHEL IUOE ClAOOfUC 1H I CKN(SS l

0.01 IN.

ZIRCON 1Ut1 l~YOAIOt- - -->

12 WT-'1.:

URAU IUM

l. 7 in.

l STAINLESS SlC(l 80T TOM EJIU f I TT I NC

~ '

Figure 4. ARRR Stainless Steel Clad Fuel Element. [13]

Page 19 of 47

Rev. 0.0 4.1.3.11 Rebar The composition data for rebar used in the activation calculations is given in Table 15.

Although no rebar is included in the flux computations, the fluxes in the bioshield will be used to estimate the activation levels in rebar. The rebar compositions are based on Reference [7].

4.1.3.12 Boral The boral was assumed to be 50% aluminum and 50% B4C by weight. This is the upper limit on B4C as reported in Reference [11].

The composition of boron carbide for the activation computations was taken from the COMETOX S.R.L. technical data on boron carbide powder for nuclear applications.

The boron carbide impurity levels were then incorporated with the aluminum impurities to arrive at the boral composition used in the activation calculations. These compositions are shown in Table 16.

4.2 Design Assumptions 4.2.1 Operating Assumptions The yearly operating data (MW-hr) will be converted to irradiation times using the licensed power, 250 kW. Then the amount of time the reactor is assumed to be operated during year is the MW-hr divided by this maximum power level. The remaining number of hours in a year will be assumed to be the length of time the reactor was not operated (down time). The down time will be assumed to occur at beginning of the year and the reactor will be assumed to be operating at full-power for the remainder of the year. This is an acceptable assumption since it provides higher activation levels than operating the reactor over the entire time period at a much lower average power. This is particularly true for the longer lived radionuclides of interest in decommissioning (half lives on the order of years).

Any operation after September 2, 2011 is assumed to be for a very short period of time to perform activities to maintain the operating license. As such, those operations should not add much to the radionuclide inventory.

4.2.2 Modeling Assumptions The assumptions for this calculation include:

Activation levels in portions of the reactor can be estimated using fluxes in nearby components if needed.

When the reactor is on, the neutron fluxes are assumed to be constant for the activation calculations.

A combination of adjusting the control rod positions and the U-235 concentrations in the fuel can be used to achieve criticality for the flux computations. The adjustment in U-235 concentrations is required since the exact operating temperature of the fuel is not known. This should not affect the fluxes in the components of interest.

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Rev. 0.0 Table 15. Rebar Constituents for Activation Calculation [7]

Average Weight High Impurity Element Fraction Fraction Fe 9.800E-01 9.800E-01 Mn 9.400E-03 1.300E-02 Cu 2.980E-03 2.530E-03 Cr 1.310E-03 1.600E-03 Ni 1.070E-03 1.260E-03 Ti 7.000E-04 7.000E-04 Mo 2.180E-04 2.740E-04 V 2.010E-04 3.340E-04 Al 1.800E-04 2.960E-04 As 1.290E-04 1.620E-04 Co 1.030E-04 1.260E-04 N 7.700E-05 7.700E-05 Zn 6.700E-05 1.600E-04 Sb 4.700E-05 1.010E-04 Nb 5.000E-06 1.510E-04 Table 16. Boral Composition for Activation Calculations.

Element Weight Fraction Weight Fraction Be 5.00E-06 2.75E-06 Ba 5.00E-06 1.26E-06 C 1.08E-01 1.08E-01 Ca 2.50E-05 1.25E-05 Cd 2.50E-06 1.25E-06 Cl 5.00E-06 2.53E-06 Co 2.50E-06 1.28E-06 Cr 2.50E-05 1.30E-05 Cu 5.00E-05 2.63E-05 Fe 7.47E-04 7.22E-04 Ga 1.00E-04 5.25E-05 Li 5.00E-06 2.75E-06 Mg 2.50E-05 1.38E-05 Mn 2.50E-05 1.38E-05 N 2.49E-03 2.49E-03 Na 2.50E-04 1.25E-04 Ni 1.00E-05 5.25E-06 O 1.84E-04 1.59E-04 P 1.50E-05 7.75E-06 Pb 2.50E-05 1.28E-05 S 1.00E-05 5.05E-06 Si 5.50E-04 3.50E-04 Sn 1.50E-05 7.53E-06 Ti 5.00E-05 2.75E-05 V 5.00E-05 2.63E-05 Zn 1.00E-04 5.25E-05 Zr 2.00E-05 1.25E-05 Al 4.98E-01 4.99E-01 B 3.89E-01 3.33E-03 Page 21 of 47

Rev. 0.0

5. Neutron Flux Calculations MCNP5[6] was used to create a three dimensional model of the ARRR. Although this code allows for a very detailed geometrical treatment of the reactor, the nature of a Monte Carlo code for a deep penetration calculation, i.e. computing the fluxes in the tank wall and the bioshield require very strong use of variance reduction techniques. This is because the transport of thermal neutrons through the water is challenging. A general description of the ARRR model created for this work, including assumptions and model simplifications that were made, follows.

5.1 ARRR Core The ARRR core was modeled using the core maps from June 10, 2011. The core maps used are seen in Figure 5 and Figure 6. The core of the ARRR consists of one central element (A-1) surrounded by six rings of elements (B-G). For rings C through G the appropriate number of elements were equally spaced around the circumference with first element in the ring positions at 180 degrees based on the core maps. For ring B the elements were equally space around the circumference with the ring rotated 18 degrees counterclockwise from the outer rings, putting element B-1 at 198 degrees based on the core maps. These rings were filled with three types of modeled elements, aluminum fuel elements, stainless steel fuel elements, and graphite fuel elements as detailed by the core map. The three control rods: safe, shim and regulating, were modeled as shown in the core map. The neutron source element was modeled as a graphite element and the instrumented aluminum fuel elements were modeled as aluminum fuel elements.

The aluminum fuel elements were modeled based primarily on details provided in the updated safety analysis report (USAR), Rev 0 for the ARRR. Figure 3 shows the dimensions of the aluminum fuel element with a cladding thickness of 0.030 inches. For the bottom end fixture and top end fixture and average diameter of the end fixture was calculated off of engineering drawings.

Stainless steel fuel element was modeled based off of details from the USAR as well. Figure 4 shows the dimensions of the fuel element with a cladding thickness of 0.020 inches. The bottom end and top end fixtures were modeled with the same diameter dimensions as the aluminum element The graphite elements were modeled with the exact same dimensions as the aluminum fuel elements, but with fuel region and the top and bottom burnable poison regions filled with graphite.

The control rods were modeled as 20 inch long rods with and outside diameter of 7/8 inches for the regulating rod and 1.25 inches for the shim and safety rods. The aluminum control rod guide tubes were modeled with an outside diameter of 1.495 inches with a cladding thickness of 0.030 inches. To account for the large number of holes evenly distribute over the entire length of the tube the density of the guide tubes was modeled as 1.7 g/cc, 2/3 the typical density of aluminum.

5.2 ARRR Criticality To correctly model the ARRR and determine the neutron fluxes throughout the three dimensional model the following adjustments were made to the model to have a critical system.

The three control rods were modeled as a high of 7.5 inches above the bottom of the fuel region in the fuel elements representing approximately half the possible travel region. The temperature of all materials in the MCNP5 model of the ARRR was kept at 293 K, or typical room Page 22 of 47

Rev. 0.0 temperature. The fuel enrichment was adjusted equally in the aluminum and stainless steel fuel elements until a critical condition was reached. The final fuel enrichment for the critical model was 12.5%.

5.3 Experimental Facilities Experimental Facilities included in the MCNP5 model of the ARRR include the Neutron Radiography Facility, Graphite Thermal Column and the Beam Port.

The Neutron Radiography Facility was modeled after the text description in the USAR for the ARRR with an 8 x 10 cross section at the bottom of the beam tube and a 22 by 34 cross section at the top of the beam tube. The bottom 48 of the beam tube was modeled with graphite with the upper section modeled with Helium. The lower 84 of the beam tube was surrounded with 3 of lead on the core side and 1 of lead on the other three sides. The distance between the core shroud and the beam tube was unknown and was modeled as 1. Details for specific structures of the neutron radiography beam tube were not available for the MCNP model and were not included. These structures include the fill and drain lines, the boral shutter mechanism.

The graphite thermal column was modeled as a 2 x 2 x 4 long graphite block encapsulated in aluminum. Specific details regarding location of the five rows of irradiation holes where not available and these holes were not modeled in the structure. Details regarding the support structure of the assembly were not available as well and were not included in the model. One may estimate the flux and therefore the activation products at the locations of the support structures based on the mesh tally data.

The beam port was modeled as a 24 outside diameter aluminum pipe ending at the aluminum reactor tank. The center of the beam port was located vertically at core centerline and extended with its axis looking directly at the center of the reactor tank. No details were available as to the material located in the beam port tube. For the purpose of modeling it was filled with air.

5.4 Core Shroud/Structure The core of the ARRR is supported by a core shroud assembly located on top of 4 T-beam one on each side running the length of the reactor pool. These T-beams were modeled as 1/2 of a standard 7 aluminum I-beam. (3.5 tall with a 4.5 wide flange, with each component being 0.23 thick) [18]. The outer shroud of the assembly is modeled as a 23.5 outer diameter aluminum cylinder of 3/4 aluminum. Within this outer shroud is a 3/4 thick bottom plate in which the fuel elements rested. This bottom plate is located 22 off the bottom of the reactor pool. Without additional details regarding the size and location of the large holes in both the lower portion of the shroud and the shroud portion surrounding the core the shroud was modeled as a solid piece of aluminum and fluxes tallied over the entire piece. Details also were not present in documentation to model any support members present between the lower shroud and the bottom plate and were not included in the MCNP model. Due to the varying heights of the core elements and limitations with the MCNP code the upper core plate was not modeled. To calculate the neutron flux at this position a mesh tally was used in which a single mesh covered the approximate location of upper core plate.

On the outside of the core shroud four water filled instruments tubes were modeled. The instrument tubes were modeled as an aluminum tube 5 in outer diameter with a 0.188 wall thickness. These tubes were extended ~10 (300 cm) from the bottom of the reactor pool to Page 23 of 47

Rev. 0.0 Figure 5. June 10, 2011 Core Map Indicating Location of Various Elements. [14]

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Rev. 0.0 Figure 6. June 10, 2011 Core Map fudicating Location of Stainless Steel Elements. [14]

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Rev. 0.0 approximately one-half way up the pool. The flux tallies on these tubes are measured the neutron flux for this entire length.

5.5 Reactor pool and Surrounding Structures The core and experimental structures were modeled inside a 10 diameter by 22 deep reactor pool that was surrounded by a 1/4 thick aluminum liner located with a 1 thick concrete biological shield. The pools and shield assembly was modeled within a ~24 thick layer of soil on the bottom and sides.

6. Activation Product Calculations 6.1 ORIGEN-S Activation and Decay Calculations This section will address the material composition input, the irradiation scenario, and the output activities for the materials of interest. Since the fluxes generated in an MCNP5 run are the fluxes for a source strength of 1 neutron in the core, the fluxes must be multiplied by the number of neutrons per second in the core. If the ARRR is operating at 250 kW, the neutron source is 3.29 1010 W fissions neutrons S 250 kW 103 2.43 kW W sec fission S 2.0 1016 neutrons sec 6.1.1 ORIGEN-S ORIGEN-S is a very flexible code and calculates activation on a user-defined basis.[1] Since the desired output of the activation and decay calculation is the radioactive material concentrations in activity per unit mass, the basis used in this calculation was per gram.

All of the ORIGEN-S activation calculations are executed by supplying the code with the number of gram-atoms of each element present in one gram of the material for the activation calculation. The ORIGEN-S output activation product activities are then output in Ci/g.

Fluxes computed using MCNP5 were used to perform the neutron activation calculations.

ORIGEN-S requires the total thermal fluxes and the three spectral parameters (FAST, RES and THERM) previously discussed to perform the activation calculation. These values were computed from the MCNP5 results. In addition the power history of the reactor must be used to create a representative neutron irradiation scenario.

6.1.1.1 Material Composition Input The material composition data needed for the activation calculations must be in gram-atoms of the element per the gram of material being activated. To calculate the gram-atoms/gram of each element in a material, the weight fractions were divided by the elemental atomic masses.

6.1.1.2 ORIGEN-S Input The ORIGEN-S input files used in the activation calculations for a given material will be identical except for the thermal flux level, the spectral constants (THERM, RES, and FAST), and the material compositions. The total thermal fluxes for each activation calculation were determined by summing the appropriate subset of the multigroup fluxes. The THERM parameter was calculated as per Equation 3.1 using the logarithmic average of the energy group boundaries Page 26 of 47

Rev. 0.0 ln Eupper ln Elower to compute 1 , where Eavg exp . The thermal flux is the sum of the E 2 group fluxes above 0.625 eV. The spectral parameter RES was calculated from Equation 3.2 using the sum of the multigroup fluxes over the energy range of 0.625 eV to 1 MeV. Similarly, FAST was computed by using the sum of the group fluxes above 1 MeV as per Equation 3.3.

6.1.2 Irradiation Scenario As previously discussed, the reported reactor operating times (kW-hr per year) were assumed to occur at maximum licensed power for each year, 250 kW. For each year, the irradiation scenario was built in this manner: The reactor was assumed to operate at full power at the end of each year for a period of time that would give the correct yearly kW-hr. For the remaining number of hours in the year, the reactor was assumed to be idle. That idle time was assigned to first part of each year. The activation calculation was performed with ORIGEN-S using the flux computed in the appropriate core regions, appropriately scaled for the maximum power level. The operating history, as used to perform the activation calculations, is given in Table 17. .

7. Activation Results Then activities were computed for 1, 5 and 10 years after October 1, 2011. Only those activities on October 1, 2012 are reported. The other activities can be extracted from the spreadsheets distributed with this report. The ORIGEN-S code runs were set up to compute the activities as a function of time for nuclides that were present on October 1, 2012 in levels higher than 10-12 curies per gram. Note that the total activities may be somewhat greater than the sum of all the radionuclide activities for a given component. This is because the code includes in its total the activities of radionuclides that were present at shutdown in levels less than 10-12 Ci/g as well.

The activation product levels are presented for representative locations in the ARRR. Some of the fluxes, and therefore the activation product levels, were calculated using the entire volume of certain objects (volume-averaged fluxes) and others were computed in localized, smaller volumes using a mesh. The volumes (cells in MCNP5) are shown in Figure 7 and Figure 8. The localized (or mesh) fluxes were computed at the locations as shown in Figure 9 and Figure 10.

These locations are further described in Table 18.

7.1 Core Support Structure.

7.1.1 Aluminum Activities The core support structure consists of the top and bottom grid plates, the cylindrical core shroud, and the shroud support. The support structure is bolted on T-beams. All this structural material is aluminum. Most of the core support structure can be seen in Figure 11. In Table 19, the activities in the Core Shroud, Top Grid Plate, Bottom Grid Plate, and Core Shroud Support Stand are shown for both the average and high impurity aluminum compositions previously defined in 4.1.3.3. The principal isotope by activity is tritium; however, no attempt has been made to correct for the diffusion of tritium out of the structures. Undoubtedly this has occurred to some degree. Also the Ar isotopes presented in that table and in other tables in this document should be probably be ignored as Ar is a noble gas and has probably diffused out of the structure. The activities in the T-beams to which the shroud support structure is bolted are shown in Table 20.

Again it is likely that the tritium has diffused out of these structures at least to some measure.

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Rev. 0.0 Table 17. Irradiation Scenario Based on Reactor Operating History Period Period Annual Full Power Years Full Power Starting Ending kW hours Off Years Operating Jun-65 Jun-66 38,152 0.9826 0.0174 Jun-66 Jun-67 57,949 0.9735 0.0265 Jun-67 Jun-68 80,409 0.9633 0.0367 Jun-68 Jun-69 61,744 0.9718 0.0282 Jun-69 Jun-70 140,279 0.9359 0.0641 Jun-70 Jun-71 215,123 0.9018 0.0982 Jun-71 Jun-72 208,086 0.9050 0.0950 Jun-72 Jun-73 196,161 0.9104 0.0896 Jun-73 Jun-74 165,159 0.9246 0.0754 Jun-74 Jun-75 203,827 0.9069 0.0931 Jun-75 Jun-76 201,868 0.9078 0.0922 Jun-76 Jun-77 244,329 0.8884 0.1116 Jun-77 Jun-78 266,586 0.8783 0.1217 Jun-78 Jun-79 293,314 0.8661 0.1339 Jun-79 Jun-80 289,969 0.8676 0.1324 Jun-80 Jun-81 281,811 0.8713 0.1287 Jun-81 Jun-82 275,699 0.8741 0.1259 Jun-82 Jun-83 309,172 0.8588 0.1412 Jun-83 Jun-84 374,479 0.8290 0.1710 Jun-84 Jun-85 445,421 0.7966 0.2034 Jun-85 Jun-86 484,242 0.7789 0.2211 Jun-86 Jun-87 487,412 0.7774 0.2226 Jun-87 Jun-88 549,864 0.7489 0.2511 Jun-88 Jun-89 653,637 0.7015 0.2985 Jun-89 Jun-90 724,951 0.6690 0.3310 Jun-90 Jun-91 690,150 0.6849 0.3151 Jun-91 Jun-92 676,144 0.6913 0.3087 Jun-92 Jun-93 371,146 0.8305 0.1695 Jun-93 Jun-94 344,002 0.8429 0.1571 Jun-94 Jun-95 350,692 0.8399 0.1601 Jun-95 Jun-96 352,727 0.8389 0.1611 Jun-96 Jun-97 340,389 0.8446 0.1554 Jun-97 Jun-98 447,745 0.7956 0.2044 Jun-98 Jun-99 369,161 0.8314 0.1686 Jun-99 Jun-00 275,791 0.8741 0.1259 Jun-00 Jun-01 240,255 0.8903 0.1097 Jun-01 Jun-02 211,950 0.9032 0.0968 Jun-02 Jun-03 222,892 0.8982 0.1018 Jun-03 Jun-04 315,652 0.8559 0.1441 Jun-04 Jun-05 284,958 0.8699 0.1301 Jun-05 Jun-06 216,433 0.9012 0.0988 Jun-06 Jun-07 252,588 0.8847 0.1153 Jun-07 Jun-08 326,056 0.8511 0.1489 Jun-08 Jun-09 279,181 0.8725 0.1275 Jun-09 Jun-10 290,319 0.8674 0.1326 Jun-10 15-Oct-10 88154 0.2529 0.0403 15-Oct-10 2-Sep-11 520 0.8759 0.0002 2-Sep-11 1-Oct-11 0 0.0831 Page 28 of 47

Rev. 0.0 Figure 7. Vertical Cut of the ARRR MCN5 Model Indicating Where Volume-Averaged Fluxes Were Computed.

Figure 8. Horizontal Cut of the ARRR MCN5 Model Indicating Where Volume-Averaged Fluxes Were Computed.

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ev. 0.0 Figme 9. Horizontal Cut Showing Locations at Which Localized Fluxes Were Computed.

Figure 10. Ve1iical Cut Showing Locations at Which Localized Fluxes Were Computed.

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Rev. 0.0 Table 18. Description of Locations in Figure 9 and Figure 10.

Location Object Material Description 1 Thermal Column Al Face of Thermal Column 1" from Core 2 Thermal Column Graphite Front of Thermal Column 1" from Core 3 Under Core Al Al reactor tank at centerline 4 Under Core Concrete 0-4" deep Concrete under core centerline 5 Under Core Concrete 8-12" deep concrete under core centerline 6 Under Core Soil 0-6" deep Soil under core centerline 7 Side wall nearest core Al Al reactor tank at centerline 8 Side wall nearest core Concrete 0-4" deep concrete at centerline 9 Side wall nearest core Concrete 8-12" deep concrete at centerline 10 Side wall nearest core Soil 0-6" deep at core centerline 11 Neutron radiography Assembly Pb front 3" lead layer at core centerline 12 Neutron radiography Assembly Al Front Al layer between lead and graphite at core centerline 13 Neutron radiography Assembly Graphite Front of Graphite at core centerline 14 Neutron radiography Assembly Graphite Back Graphite at core centerline 15 Neutron radiography Assembly AL Back Al layer between lead and graphite at core centerline 16 Neutron radiography Assembly Pb back 1" lead at core centerline 17 Neutron radiography Shutter Boral Shutter at top of graphine block in vertical beam tube 18 Beam Port Concrete Concrete nearest core at beam port entrance 19 Side wall behind thermal column Al Al reactor tank at centerline 20 Side wall behind thermal column Concrete 0-4" deep concrete at centerline Figure 11. ARRR Core and Support Structure.

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Rev. 0.0 Table 19. Activation Levels as of October 1, 2012 in the Core Shroud, Top Grid Plate, Bottom Grid Plate, and Core Shroud Support Stand in Ci/g.

Bottom Grid Plate Top Grid Plate Core Shroud Core Support Shroud Radionuclide Average Impurity High Impurity Average Impurity High Impurity Average Impurity High Impurity Average Impurity High Impurity H-3 2.07E-05 3.77E-05 4.09E-05 7.43E-05 7.45E-05 1.36E-04 1.84E-06 3.35E-06 C-14 1.09E-09 1.91E-09 2.10E-09 3.68E-09 3.70E-09 6.47E-09 1.00E-10 1.75E-10 Na-22 2.44E-12 4.87E-12 6.43E-12 1.29E-11 8.59E-12 1.72E-11 S-35 1.83E-10 3.62E-10 3.79E-10 7.51E-10 6.41E-10 1.27E-09 1.60E-11 3.16E-11 Cl-36 1.87E-10 3.70E-10 3.58E-10 7.09E-10 6.30E-10 1.25E-09 1.71E-11 3.39E-11 Ar-37 7.19E-12 1.44E-11 1.86E-11 3.70E-11 2.81E-11 5.62E-11 1.03E-12 Ar-39 1.04E-12 Ca-41 2.89E-11 5.78E-11 5.56E-11 1.11E-10 9.78E-11 1.95E-10 2.65E-12 5.29E-12 Ca-45 1.55E-09 3.10E-09 2.98E-09 5.96E-09 5.25E-09 1.05E-08 1.42E-10 2.84E-10 Sc-46 8.92E-12 1.63E-11 2.55E-11 4.67E-11 4.02E-11 7.40E-11 Cr-51 7.05E-11 1.36E-10 1.35E-10 2.60E-10 2.38E-10 4.58E-10 6.46E-12 1.24E-11 Mn-54 1.03E-09 1.15E-09 2.89E-09 3.22E-09 4.23E-09 4.72E-09 6.63E-11 7.38E-11 Fe-55 3.21E-07 3.57E-07 6.17E-07 6.85E-07 1.09E-06 1.21E-06 2.94E-08 3.27E-08 Fe-59 1.35E-10 1.50E-10 2.62E-10 2.91E-10 4.59E-10 5.10E-10 1.23E-11 1.37E-11 Co-58 4.11E-11 7.82E-11 1.15E-10 2.20E-10 1.69E-10 3.21E-10 2.65E-12 5.05E-12 Co-60 5.95E-07 1.17E-06 1.16E-06 2.28E-06 2.02E-06 3.96E-06 5.41E-08 1.06E-07 Ni-59 9.05E-11 1.72E-10 1.73E-10 3.30E-10 3.05E-10 5.80E-10 8.30E-12 1.58E-11 Ni-63 9.78E-09 1.86E-08 1.89E-08 3.60E-08 3.31E-08 6.31E-08 8.93E-10 1.70E-09 Zn-65 1.21E-07 2.30E-07 2.37E-07 4.51E-07 4.12E-07 7.84E-07 1.10E-08 2.09E-08 Zr-95 6.63E-11 1.06E-10 1.37E-10 2.19E-10 2.32E-10 3.71E-10 5.86E-12 9.38E-12 Nb-95 1.46E-10 2.34E-10 3.02E-10 4.83E-10 5.11E-10 8.18E-10 1.29E-11 2.07E-11 Nb-95m 1.25E-12 1.61E-12 2.58E-12 2.73E-12 4.36E-12 Ag-109m 9.41E-11 1.88E-10 2.03E-10 4.06E-10 3.35E-10 6.69E-10 8.14E-12 1.62E-11 Cd-109 9.41E-11 1.88E-10 2.03E-10 4.06E-10 3.35E-10 6.69E-10 8.14E-12 1.62E-11 Cd-113m 9.45E-10 1.89E-09 1.36E-09 2.71E-09 1.59E-09 3.17E-09 1.34E-10 2.68E-10 Cd-115m 4.00E-12 7.98E-12 9.52E-12 1.90E-11 1.55E-11 3.09E-11 In-113m 2.03E-10 4.04E-10 4.63E-10 9.23E-10 7.43E-10 1.48E-09 1.70E-11 3.38E-11 Sn-113 2.03E-10 4.04E-10 4.63E-10 9.22E-10 7.43E-10 1.48E-09 1.70E-11 3.38E-11 Sn-119m 6.15E-09 1.23E-08 1.46E-08 2.91E-08 2.30E-08 4.58E-08 5.02E-10 1.00E-09 Sn-121 7.67E-12 1.53E-11 1.60E-11 3.20E-11 2.70E-11 5.37E-11 1.34E-12 Sn-121m 9.88E-12 1.97E-11 2.07E-11 4.12E-11 3.47E-11 6.92E-11 1.73E-12 Sn-123 8.93E-11 1.78E-10 1.79E-10 3.57E-10 3.08E-10 6.14E-10 8.01E-12 1.60E-11 Sb-125 8.82E-10 1.76E-09 2.04E-09 4.07E-09 3.26E-09 6.49E-09 7.32E-11 1.46E-10 Te-125m 2.15E-10 4.28E-10 4.98E-10 9.92E-10 7.92E-10 1.58E-09 1.78E-11 3.55E-11 Cs-134 1.95E-12 2.88E-12 5.71E-12 Ba-133 3.79E-10 7.50E-10 7.39E-10 1.46E-09 1.29E-09 2.55E-09 3.45E-11 6.82E-11 Total 2.18E-05 3.95E-05 4.29E-05 7.78E-05 7.81E-05 1.42E-04 1.94E-06 3.52E-06 Page 32 of 47

Rev. 0.0 Table 20. T-Beam Activities (Ci/g) on October 1, 2012. See Figure 7 for the locations of the cells.

Cell 248 Cell 245 Cell 254 Cell 251 Radionuclide Average Impurity High Impurity Average Impurity High Impurity Average Impurity High Impurity Average Impurity High Impurity H-3 1.48E-08 2.70E-08 3.45E-08 6.28E-08 7.13E-09 1.30E-08 1.48E-08 2.70E-08 C-14 1.41E-12 1.88E-12 3.28E-12 1.41E-12 Ca-45 1.15E-12 2.29E-12 2.67E-12 5.33E-12 1.10E-12 1.15E-12 2.29E-12 Mn-54 1.07E-12 1.20E-12 Fe-55 2.37E-10 2.63E-10 5.51E-10 6.13E-10 1.14E-10 1.27E-10 2.37E-10 2.63E-10 Co-60 4.35E-10 8.52E-10 1.01E-09 1.98E-09 2.10E-10 4.12E-10 4.34E-10 8.50E-10 Ni-63 7.24E-12 1.38E-11 1.67E-11 3.19E-11 3.50E-12 6.66E-12 7.21E-12 1.37E-11 Zn-65 8.80E-11 1.68E-10 2.04E-10 3.88E-10 4.26E-11 8.11E-11 8.77E-11 1.67E-10 Cd-113m 1.13E-12 2.26E-12 2.60E-12 5.19E-12 1.11E-12 1.12E-12 2.24E-12 Sn-119m 3.82E-12 7.62E-12 8.36E-12 1.67E-11 2.02E-12 4.03E-12 3.66E-12 7.29E-12 Sb-125 1.11E-12 1.24E-12 2.47E-12 1.07E-12 Ba-133 1.28E-12 Total 1.56E-08 2.83E-08 3.63E-08 6.58E-08 7.50E-09 1.36E-08 1.56E-08 2.83E-08 Page 33 of 47

Rev. 0.0 7.1.2 Bolts in T-Bars Activities Activation calculations were performed for the bolts in the T-bars (Cells 248 and 245). The bolts were assumed to be SS304 (see Section4.1.3.4). The activities are in Table 21.

7.2 Instrument Guide Tubes The activities for the instrument guide tubes are reported in Table 22. These activities were computed using the volume-averaged fluxes in the tubes over the first 300 cm of their length, starting on the bottom of the core support structure.

7.3 Thermal Column and Neutron Radiography Assembly 7.3.1 Graphite Activities The activities in the thermal column and neutron radiography assembly were computed for the volume-averaged flux in the two graphite blocks in those assemblies (cell 215 and 220, respectively), using the flux in the first 1 of the graphite for the thermal column (nearest the core, location 2) and using the flux in the first and last 1 of graphite for the neutron radiography assembly (locations 13 and 14). The localized fluxes used were for small volumes on the centerline of the core, so they represent close to the maximum fluxes at those locations. Only one impurity composition for graphite was used (Section 4.1.3.7). For completeness, the activities induced by adding 1 ppm Eu and 1 ppm Cl by weight have been added to the tables. In some nuclear grade graphite, either of those two elements may be present depending on the purification method. However, it is not known what the specifications on the graphite were.

Again the Ar isotopes have probably migrated out of the assemblies.

7.3.2 Aluminum Activities The activities in the aluminum in the thermal column and the radiography assembly were also computed. Activities were computed using the average flux in the thermal column encapsulation (cell 221), local flux at the front face of the thermal column (location 1), and the local fluxes at the front and back of the neutron radiography assembly (locations 12 and 15). The localized fluxes again represent nearly the maximum flux at that depth in the assemblies. These activities are reported in Table 24. Again keep in mind that the tritium has at least partially diffused out of these components and as has the argon.

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Rev. 0.0 Table 21. Activities (Ci/g) in Representative Bolts in the T-Beams of the Structure on October 1, 2012. The bolts were assumed to be SS304.

Cell 248 Cell 245 Radionuclide Average Impurities High Impurities Average Impurities High Impurities C-14 8.98E-11 1.06E-10 2.09E-10 2.46E-10 Cr-51 3.64E-10 3.59E-10 8.46E-10 8.34E-10 Mn-54 1.17E-09 1.15E-09 1.66E-09 1.64E-09 Fe-55 3.68E-07 3.62E-07 8.55E-07 8.42E-07 Fe-59 1.54E-10 1.51E-10 3.57E-10 3.51E-10 Co-58 2.83E-10 2.80E-10 4.03E-10 3.98E-10 Co-60 2.37E-07 4.38E-07 5.50E-07 1.02E-06 Ni-59 6.29E-10 6.22E-10 1.46E-09 1.45E-09 Ni-63 6.69E-08 6.62E-08 1.56E-07 1.54E-07 Zn-65 3.78E-10 1.87E-09 8.74E-10 4.33E-09 Nb-93m 0.00E+00 0.00E+00 0.00E+00 1.18E-12 Nb-94 0.00E+00 2.10E-12 1.40E-12 4.80E-12 Mo-93 0.00E+00 1.20E-12 1.22E-12 2.62E-12 Ce-139 0.00E+00 0.00E+00 1.48E-12 2.22E-12 Ce-141 0.00E+00 1.07E-12 1.64E-12 2.48E-12 Eu-152 8.34E-11 8.47E-11 1.94E-10 1.97E-10 Eu-154 6.02E-12 6.11E-12 1.39E-11 1.41E-11 W-181 3.82E-12 1.08E-11 8.77E-12 2.49E-11 W-185 1.76E-11 4.99E-11 4.02E-11 1.14E-10 Total 6.75E-07 8.71E-07 1.57E-06 2.02E-06 Page 35 of 47

Rev. 0.0 Table 22. Activities (Ci/g) in the Instrument Guide Tubes on October 1, 2012.

Instrument Tube 1 Instrument Tube 2 Instrument Tube 3 Instrument Tube 4 Radionuclide Average Impurity High Impurity Average Impurity High Impurity Average Impurity High Impurity Average Impurity High Impurity H-3 8.15E-06 1.48E-05 7.12E-06 1.30E-05 6.87E-06 1.25E-05 6.03E-06 1.10E-05 C-14 4.37E-10 7.66E-10 3.83E-10 6.70E-10 3.70E-10 6.47E-10 3.25E-10 5.69E-10 Na-22 1.89E-12 1.62E-12 1.61E-12 1.35E-12 S-35 7.24E-11 1.43E-10 6.31E-11 1.25E-10 6.12E-11 1.21E-10 5.34E-11 1.06E-10 Cl-36 7.49E-11 1.48E-10 6.55E-11 1.30E-10 6.33E-11 1.25E-10 5.56E-11 1.10E-10 Ar-37 2.73E-12 5.45E-12 2.34E-12 4.67E-12 2.31E-12 4.61E-12 1.96E-12 3.91E-12 Ca-41 1.16E-11 2.31E-11 1.01E-11 2.03E-11 9.80E-12 1.96E-11 8.61E-12 1.72E-11 Ca-45 6.22E-10 1.24E-09 5.45E-10 1.09E-09 5.26E-10 1.05E-09 4.62E-10 9.23E-10 Sc-46 3.22E-12 5.87E-12 2.74E-12 4.99E-12 2.72E-12 4.95E-12 2.27E-12 4.14E-12 Cr-51 2.82E-11 5.43E-11 2.47E-11 4.75E-11 2.39E-11 4.59E-11 2.10E-11 4.03E-11 Mn-54 3.83E-10 4.26E-10 3.26E-10 3.63E-10 3.23E-10 3.60E-10 2.71E-10 3.02E-10 Fe-55 1.29E-07 1.43E-07 1.13E-07 1.25E-07 1.09E-07 1.21E-07 9.55E-08 1.06E-07 Fe-59 5.40E-11 6.00E-11 4.72E-11 5.25E-11 4.56E-11 5.07E-11 4.01E-11 4.45E-11 Co-58 1.53E-11 2.91E-11 1.30E-11 2.48E-11 1.29E-11 2.46E-11 1.08E-11 2.06E-11 Co-60 2.38E-07 4.66E-07 2.08E-07 4.08E-07 2.01E-07 3.94E-07 1.76E-07 3.46E-07 Ni-59 3.63E-11 6.91E-11 3.18E-11 6.05E-11 3.07E-11 5.84E-11 2.70E-11 5.13E-11 Ni-63 3.91E-09 7.46E-09 3.43E-09 6.53E-09 3.31E-09 6.30E-09 2.91E-09 5.54E-09 Zn-65 4.82E-08 9.18E-08 4.22E-08 8.03E-08 4.07E-08 7.76E-08 3.58E-08 6.81E-08 Zr-95 2.63E-11 4.20E-11 2.29E-11 3.66E-11 2.22E-11 3.55E-11 1.94E-11 3.10E-11 Nb-95 5.79E-11 9.27E-11 5.05E-11 8.08E-11 4.89E-11 7.83E-11 4.28E-11 6.84E-11 Ag-109m 3.70E-11 7.39E-11 3.22E-11 6.43E-11 3.13E-11 6.24E-11 2.72E-11 5.44E-11 Cd-109 3.70E-11 7.39E-11 3.22E-11 6.43E-11 3.13E-11 6.24E-11 2.72E-11 5.44E-11 Cd-113m 5.01E-10 1.00E-09 4.49E-10 8.97E-10 4.38E-10 8.74E-10 3.92E-10 7.83E-10 Cd-115m 1.40E-12 2.79E-12 1.19E-12 2.38E-12 1.16E-12 2.31E-12 1.97E-12 In-113m 7.90E-11 1.57E-10 6.84E-11 1.36E-10 6.66E-11 1.33E-10 5.78E-11 1.15E-10 Sn-113 7.89E-11 1.57E-10 6.83E-11 1.36E-10 6.65E-11 1.33E-10 5.77E-11 1.15E-10 Sn-119m 2.38E-09 4.74E-09 2.05E-09 4.09E-09 2.00E-09 3.99E-09 1.73E-09 3.45E-09 Sn-121 3.03E-12 6.04E-12 2.64E-12 5.26E-12 2.56E-12 5.10E-12 2.24E-12 4.46E-12 Sn-121m 3.91E-12 7.78E-12 3.40E-12 6.78E-12 3.30E-12 6.57E-12 2.88E-12 5.74E-12 Sn-123 3.55E-11 7.08E-11 3.10E-11 6.18E-11 3.00E-11 5.98E-11 2.63E-11 5.24E-11 Sb-125 3.43E-10 6.83E-10 2.96E-10 5.90E-10 2.89E-10 5.75E-10 2.50E-10 4.99E-10 Te-125m 8.34E-11 1.66E-10 7.21E-11 1.44E-10 7.03E-11 1.40E-10 6.09E-11 1.21E-10 Ba-133 1.51E-10 3.00E-10 1.32E-10 2.62E-10 1.28E-10 2.53E-10 1.12E-10 2.23E-10 Total 8.57E-06 1.55E-05 7.49E-06 1.36E-05 7.23E-06 1.31E-05 6.35E-06 1.15E-05 Page 36 of 47

Rev. 0.0 Table 23. Activities in the Graphite (Ci/g) in the Thermal Column and Neutron Radiography Assembly on October 1, 2012.

Thermal Column Neutron Radiography Assembly Radionuclide Entire Block Front Face Entire Block Front Face Back Face Be-10 6.19E-12 4.44E-12 C-14 8.48E-10 9.04E-09 1.04E-09 5.00E-09 5.07E-10 S-35 7.93E-16 8.74E-10 9.44E-16 4.68E-10 5.06E-11 Ar-37 1.87E-24 1.30E-10 2.96E-24 8.74E-11 2.57E-12 Ar-39 1.87E-12 1.36E-12 Ca-41 5.11E-11 5.31E-10 6.09E-11 2.85E-10 3.07E-11 Ca-45 5.66E-12 2.85E-08 6.73E-12 1.53E-08 1.65E-09 Sc-46 3.19E-11 1.14E-11 Mn-54 9.56E-14 7.02E-11 1.88E-13 5.03E-11 Fe-55 7.77E-10 2.23E-08 9.26E-10 1.20E-08 1.29E-09 Fe-59 9.47E-12 1.42E-22 5.15E-12 Co-60 2.30E-12 Total 1.68E-09 6.15E-08 2.03E-09 3.31E-08 3.53E-09 1 ppm Europium Eu-152 1.29E-06 7.58E-06 1.50E-06 6.03E-06 9.84E-07 Eu-154 8.98E-08 1.28E-06 1.12E-07 7.39E-07 7.41E-08 Eu-155 6.23E-10 1.08E-07 9.25E-10 3.53E-08 4.07E-10 Gd-153 7.05E-11 3.19E-07 9.87E-11 1.15E-07 1.72E-09 1 ppm Chlorine S-35 2.23E-11 1.74E-17 1.61E-11 Cl-36 1.12E-11 1.15E-10 1.33E-11 6.18E-11 6.71E-12 Page 37 of 47

Rev. 0.0 Table 24. Activities (Ci/g) in the Aluminum in Representative Locations of the Thermal Column and the Neutron Radiography Assembly on October 1, 2012.

Thermal Column Assembly Neutron Radiography Assembly Radionuclide Aluminum Encapsulation Front Face at Core Centerline Front Aluminum Layer Back Aluminum Layer Average Impurities High Impurities Average Impurities High Impurities Average Impurities High Impurities Average Impurities High Impurities H-3 5.40E-06 9.83E-06 6.85E-05 1.26E-04 4.71E-05 8.56E-05 3.66E-06 6.65E-06 C-14 2.91E-10 5.10E-10 3.42E-09 6.03E-09 2.41E-09 4.22E-09 1.97E-10 3.45E-10 Na-22 6.54E-12 1.30E-11 1.01E-11 2.02E-11 S-35 4.63E-11 9.17E-11 5.77E-10 1.15E-09 4.77E-10 9.44E-10 2.87E-11 5.67E-11 Cl-36 4.99E-11 9.88E-11 5.82E-10 1.16E-09 4.09E-10 8.10E-10 3.39E-11 6.71E-11 Ar-37 1.44E-12 2.88E-12 2.22E-11 4.43E-11 2.81E-11 5.61E-11 Ar-39 1.02E-12 Ca-41 7.72E-12 1.54E-11 9.05E-11 1.82E-10 6.37E-11 1.27E-10 5.23E-12 1.05E-11 Ca-45 4.14E-10 8.27E-10 4.86E-09 9.79E-09 3.42E-09 6.83E-09 2.81E-10 5.61E-10 Sc-46 1.54E-12 2.81E-12 3.07E-11 5.67E-11 4.04E-11 7.38E-11 Cr-51 1.88E-11 3.62E-11 2.20E-10 4.27E-10 1.55E-10 2.98E-10 1.28E-11 2.45E-11 Mn-54 1.83E-10 2.04E-10 3.17E-09 3.53E-09 4.63E-09 5.16E-09 2.06E-11 2.30E-11 Fe-55 8.56E-08 9.51E-08 1.00E-06 1.13E-06 7.07E-07 7.86E-07 5.80E-08 6.45E-08 Fe-59 3.59E-11 3.99E-11 4.26E-10 4.77E-10 3.05E-10 3.39E-10 2.43E-11 2.70E-11 Co-58 7.32E-12 1.40E-11 1.26E-10 2.41E-10 1.85E-10 3.52E-10 1.57E-12 Co-60 1.58E-07 3.10E-07 1.88E-06 3.72E-06 1.37E-06 2.68E-06 1.07E-07 2.09E-07 Ni-59 2.42E-11 4.60E-11 2.82E-10 5.42E-10 1.98E-10 3.78E-10 1.64E-11 3.12E-11 Ni-63 2.60E-09 4.95E-09 3.05E-08 5.86E-08 2.18E-08 4.15E-08 1.75E-09 3.33E-09 Zn-65 3.20E-08 6.09E-08 3.83E-07 7.35E-07 2.80E-07 5.33E-07 2.15E-08 4.10E-08 Zr-95 1.72E-11 2.75E-11 2.19E-10 3.50E-10 1.75E-10 2.80E-10 1.13E-11 1.81E-11 Nb-95 3.80E-11 6.08E-11 4.82E-10 7.73E-10 3.86E-10 6.17E-10 2.49E-11 3.98E-11 Nb-95m 2.57E-12 4.12E-12 2.06E-12 3.29E-12 Ag-109m 2.41E-11 4.81E-11 3.23E-10 6.42E-10 2.76E-10 5.50E-10 1.55E-11 3.10E-11 Cd-109 2.41E-11 4.81E-11 3.23E-10 6.42E-10 2.76E-10 5.50E-10 1.55E-11 3.10E-11 Cd-113m 3.56E-10 7.10E-10 1.58E-09 3.15E-09 1.55E-09 3.10E-09 2.48E-10 4.96E-10 Cd-115m 1.71E-12 1.52E-11 3.00E-11 1.39E-11 2.77E-11 1.04E-12 In-113m 5.07E-11 1.01E-10 7.29E-10 1.44E-09 6.72E-10 1.34E-09 3.19E-11 6.35E-11 Sn-113 5.07E-11 1.01E-10 7.29E-10 1.44E-09 6.71E-10 1.34E-09 3.18E-11 6.34E-11 Sn-119m 1.51E-09 3.01E-09 2.27E-08 4.47E-08 2.20E-08 4.39E-08 9.22E-10 1.84E-09 Sn-121 1.98E-12 3.95E-12 2.56E-11 5.11E-11 2.09E-11 4.17E-11 1.30E-12 2.58E-12 Sn-121m 2.56E-12 5.09E-12 3.30E-11 6.58E-11 2.69E-11 5.37E-11 1.67E-12 3.33E-12 Sn-123 2.34E-11 4.67E-11 2.88E-10 5.77E-10 2.20E-10 4.38E-10 1.56E-11 3.11E-11 Sb-125 2.19E-10 4.37E-10 3.21E-09 6.33E-09 3.01E-09 6.01E-09 1.37E-10 2.73E-10 Te-125m 5.34E-11 1.07E-10 7.82E-10 1.54E-09 7.34E-10 1.46E-09 3.33E-11 6.64E-11 Cs-134 2.55E-12 5.09E-12 1.50E-12 2.96E-12 Ba-133 1.01E-10 1.99E-10 1.20E-09 2.39E-09 8.71E-10 1.72E-09 6.78E-11 1.34E-10 Total 5.69E-06 1.03E-05 7.18E-05 1.32E-04 4.95E-05 8.97E-05 3.85E-06 6.97E-06 Page 38 of 47

Rev. 0.0 7.3.3 Lead Activities Activities were also calculated in the lead layers of the radiography assembly by using localized fluxes (locations 11 and 16), 3 of the front lead layer and 1 of the back lead layer. The resulting activities are shown Table 25. Note that the total activity on October 1, 2012 is driven by long-lived metastable Ag isotopes using the assumed impurity levels (Section 4.1.3.2) 7.3.4 Boral Shutter Activities The boral shutter was not included in the model of the neutron radiography assembly used in the flux calculations, the assumption that it was fully opened during use. So the flux in the neutron radiography assembly at the location at which the shutter would have been when fully closed was used to compute the activities in the boral. The impurity levels described in Section 4.1.3.12 were used. The use of this flux will lead to higher activation levels in the boral since it provides a higher thermal flux throughout the boral than would actually exist since it is a strong thermal neutron absorber. The resulting activities are reported in Table 26.

7.4 Aluminum Tank Activities The aluminum activities in the tank were calculated using localized neutron fluxes computed in small volumes underneath the core at the core centerline and at the side of the tank nearest the core on the core centerline. These activities represent near maximum aluminum activities in the tank wall. The total activities are 1000 - 10000 times less than in the structures nearer the core.

7.5 Bioshield Activities 7.5.1 Concrete Activities The activities for the concrete bioshield are presented Table 28 for the impurity representations given in Section 4.1.3.1. These activities are computed using local flux averages in small volumes underneath the core (locations 4 and 5) and at the nearest wall (locations 8 and 9). The resulting activity concentrations are shown in Table 28.

7.5.2 Rebar Activities No rebar was included in the flux model calculations. So the fluxes at the positions used in the previous subsection were used with the rebar compositions presented in Section 4.1.3.11. The resulting activities are given in Table 29.

7.6 Soil Activities As previously mentioned, no composition was available for red clay. So an average composition for U.S. soils was used in the activation calculation (see Section 4.1.3.5). The activities in Table 30 were computed using the fluxes at locations 6 and 10. The Ar-40 may have diffused out and the K-40 is not from activation but from the potassium in the soil composition used.

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Rev. 0.0 Table 25. Activities in the Lead (Ci/g) at the Front and Back of the Neutron Radiography Assembly on October 1, 2012.

Radionuclide Front Back Mn-54 3.60E-10 Fe-55 6.01E-08 2.55E-09 Fe-59 2.57E-11 1.07E-12 Co-60 3.50E-11 Ni-63 3.27E-10 Zn-65 5.02E-08 2.03E-09 Ag-108 2.52E-08 1.02E-09 Ag-108m 2.89E-07 1.17E-08 Ag-109m 2.99E-09 3.85E-12 Ag-110 2.18E-08 7.70E-10 Ag-110m 1.60E-06 5.67E-08 Cd-109 2.99E-09 3.85E-12 In-113m 5.03E-10 1.40E-11 Sn-113 5.02E-10 1.40E-11 Sn-119m 1.62E-08 4.04E-10 Sn-121 1.65E-11 Sn-121m 2.13E-11 Sn-123 1.80E-10 6.84E-12 Sb-124 1.51E-08 4.87E-10 Sb-125 2.24E-09 6.00E-11 Te-123m 6.96E-11 Te-125m 5.45E-10 1.46E-11 Hg-203 8.34E-12 Tl-204 1.10E-10 Pb-205 1.78E-10 6.88E-12 Po-210 3.19E-08 1.16E-09 Total 2.12E-06 7.69E-08 Page 40 of 47

Rev. 0.0 Table 26. Boral Shutter Activities (Ci/g) on October 1, 2012 for Average and High Impurity Levels Radionuclide Average Impurities High Impurities H-3 1.79E-06 3.25E-06 Be-10 2.40E-11 2.40E-11 C-14 1.20E-07 1.21E-07 S-35 1.70E-11 3.37E-11 Cl-36 1.65E-11 3.28E-11 Ar-37 1.50E-12 Ca-41 2.57E-12 5.12E-12 Ca-45 1.38E-10 2.75E-10 Sc-46 1.69E-12 Cr-51 6.25E-12 1.20E-11 Mn-54 3.59E-10 3.73E-10 Fe-55 9.11E-08 9.46E-08 Fe-59 3.88E-11 4.03E-11 Co-58 4.49E-12 8.55E-12 Co-60 5.38E-08 1.06E-07 Ni-59 8.02E-12 1.53E-11 Ni-63 8.68E-10 1.65E-09 Zn-65 1.10E-08 2.09E-08 Zr-95 6.40E-12 1.02E-11 Nb-95 1.41E-11 2.26E-11 Ag-109m 9.62E-12 1.92E-11 Cd-109 9.62E-12 1.92E-11 Cd-113m 1.34E-10 2.68E-10 In-113m 2.23E-11 4.43E-11 Sn-113 2.23E-11 4.42E-11 Sn-119m 7.04E-10 1.40E-09 Sn-121 1.50E-12 Sn-121m 1.94E-12 Sn-123 8.34E-12 1.66E-11 Sb-125 9.87E-11 1.96E-10 Te-125m 2.40E-11 4.77E-11 Ba-133 1.72E-11 6.79E-11

. Total 2.06E-06 3.59E-06 Table 27. Aluminum Tank Activities (Ci/g) on October 1, 2012.

Underneath Core on Centerline Nearest Wall on Core Centerline Radionuclide Average Impurities High Impurities Average Impurities High Impurities H-3 4.04E-09 7.32E-09 8.23E-10 1.50E-09 Fe-55 6.47E-11 7.16E-11 1.32E-11 1.47E-11 Co-60 1.21E-10 2.35E-10 2.49E-11 4.88E-11 Ni-63 2.03E-12 3.86E-12 Zn-65 2.45E-11 4.65E-11 5.08E-12 9.68E-12 Sn-119m 1.36E-12 2.70E-12 Total 4.26E-09 7.68E-09 8.67E-10 1.57E-09 Page 41 of 47

Rev. 0.0 Table 28. Activities in the Concrete Below and to the Nearest Side of the Core (Ci/g) on October 1, 2012.

Underneath Core (0-4") Underneath Core (8-12") Nearest Side (0-4") Nearest Side (8-12")

Radionuclides Average Impurities High Impurities Average Impurities High Impurities Average Impurities High Impurities Average Impurities High Impurities C-14 8.09E-12 7.98E-12 2.05E-12 2.02E-12 1.65E-12 1.63E-12 Ar-37 6.52E-12 6.40E-12 1.18E-12 1.16E-12 1.35E-12 1.33E-12 Ca-41 1.02E-11 9.99E-12 2.60E-12 2.55E-12 2.07E-12 2.04E-12 Ca-45 5.46E-10 5.36E-10 1.39E-10 1.37E-10 1.11E-10 1.09E-10 2.82E-11 2.77E-11 Mn-54 1.98E-11 1.96E-11 3.42E-12 3.39E-12 4.12E-12 4.08E-12 Fe-55 1.99E-09 1.97E-09 5.08E-10 5.02E-10 4.06E-10 4.01E-10 1.03E-10 1.02E-10 Co-60 4.59E-10 1.43E-09 1.18E-10 3.67E-10 9.50E-11 2.96E-10 2.41E-11 7.49E-11 Ni-63 7.00E-12 1.59E-11 1.78E-12 4.04E-12 1.43E-12 3.23E-12 Zn-65 1.74E-11 7.76E-11 4.47E-12 1.99E-11 3.61E-12 1.61E-11 4.07E-12 Ba-133 1.44E-11 1.05E-10 3.68E-12 2.70E-11 2.97E-12 2.18E-11 5.51E-12 Eu-152 1.12E-09 3.48E-09 2.85E-10 8.87E-10 2.27E-10 7.07E-10 5.77E-11 1.80E-10 Eu-154 8.36E-11 2.60E-10 2.16E-11 6.73E-11 1.76E-11 5.48E-11 4.44E-12 1.38E-11 Total 4.28E-09 7.92E-09 1.09E-09 2.02E-09 8.74E-10 1.62E-09 2.22E-10 4.11E-10 Table 29. Activities in the Rebar Below and to the Nearest Side of the Core (Ci/g) on October 1, 2012.

Below Core (0-4") Below Core (8-12 inches) Behind Tank Wall Nearest Core (0-4") Behind Tank Wall Nearest Core (8-12")

Radionuclide Average Impurities High Impurities Average Impurities High Impurities Average Impurities High Impurities Average Impurities High Impurities C-14 4.26E-12 4.26E-12 1.09E-12 1.09E-12 Mn-54 1.40E-09 1.40E-09 2.41E-10 2.41E-10 2.90E-10 2.90E-10 6.14E-11 6.14E-11 Fe-55 1.41E-07 1.41E-07 3.59E-08 3.59E-08 2.86E-08 2.86E-08 7.28E-09 7.28E-09 Fe-59 5.92E-11 5.92E-11 1.52E-11 1.52E-11 1.22E-11 1.22E-11 3.09E-12 3.09E-12 Co-58 2.62E-12 3.10E-12 Co-60 4.87E-09 5.96E-09 1.25E-09 1.53E-09 1.01E-09 1.23E-09 2.55E-10 3.12E-10 Ni-59 1.86E-12 2.20E-12 Ni-63 2.03E-10 2.38E-10 5.13E-11 6.03E-11 4.13E-11 4.84E-11 1.04E-11 1.23E-11 Zn-65 1.56E-11 3.74E-11 4.00E-12 9.60E-12 3.23E-12 7.75E-12 1.96E-12 Sb-124 4.33E-12 9.31E-12 1.15E-12 2.47E-12 2.09E-12 Total 1.47E-07 1.48E-07 3.74E-08 3.77E-08 3.00E-08 3.02E-08 7.61E-09 7.67E-09 Page 42 of 47

Rev. 0.0 Table 30. Activity in Soil (Ci/g) on October 1, 2012.

Under the Core Behind Nearest Radionuclide (0-6") Side Wall (0-6")

Ar-39 1.02E-11 2.96E-12 K-40 1.20E-11 1.20E-11 Ca-41 1.26E-12 Ca-45 6.73E-11 1.14E-11 Mn-54 5.64E-12 1.63E-12 Fe-55 8.50E-10 1.44E-10 Total 9.48E-10 1.73E-10

8. Summary The activity product levels in components of the Aerotest Radiography and Research Reactor have been calculated using a three dimensional MCNP5 model. The activities have been presented for representative locations and can be computed for more subject to the clients wishes. The activities were based on an operating scenario in which each years burnup hours were assumed to occur at full power, 250 kW, at the end of the year. The activation levels rely on assumptions about impurity levels in the construction materials. Since no chemical analyses were available for the construction materials, representative impurity levels were derived based on various documents and books were used to form material compositions for the activation calculations. The activities of aluminum components often showed that tritium was the principal contributor to total activity. However, no attempt was made to correct the activation code results for diffusion of tritium out of the components over the operating period. Similarly, the activation calculations often indicated activity from argon isotopes. Again in this case, it is like that significant fractions if not all the argon has diffused out of the system.

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Rev. 0.0

9. References

[1] I. C. Gauld, O. W. Hermann, R. M. Westfall, ORIGEN-S: SCALE System Module to Calculate Fuel Depletion, Actinide Transmutation, Fission Product Buildup and Decay, and Associated Radiation Source Terms, Oak Ridge National Laboratory, ORNL/TM-2005/39, Volume 2, Section F7, January 2009.

[2] Oak Ridge National Laboratory, SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, ORNL/TM-2005/39, Version 6, Volumes I-III, January 2009.

[3] I. C. Gauld, B. D. Murphy, M. L. Williams, ORIGEN-S Data Libraries, Oak Ridge National Laboratory, ORNL/TM-2005/39, Volume III, Section M6, January 2009.

[4] L. M. Petrie, P. B. Fox and K. Lucius, Standard Composition Library, Oak Ridge National Laboratory, ORNL/TM-2005/39, Volume III, Section M8, January 2009.

[5] J. K. Tuli (ed.), Nuclear Wallet Cards, 6th Edition, National Nuclear Data Center, Brookhaven National Laboratory, January 2000.

[6] X-5 Monte Carlo Team. MCNP - A General Monte Carlo N-Particle Tansport Code Version 5. LA-CP-03-0245. Vol. II. (Los Alamos National Laboratory) (2004).

[7] J. C. Evans, E. L. Lepel, R. W. Sanders, C. L. Wilkerson, W. Silker, C. W. Thomas, K.

H. Abdel, and D. R. Robertson, Long-Lived Activation Products in Reactor Materials, U. S. Nuclear Regulatory Commission, NUREG/CR -3474, 1984.

[8] Michael Gauccio (ed.), ASM Metals Reference Book, 3rd Edition, ASM, 1993.

[9] G.J. Konzek, J. D. Ludwick, W. E. Kennedy, Jr., and R. I. Smith, Technology, Safety and Costs of Decommissioning Reference Research and Test Reactors, U. S. Nuclear Regulatory Commission, NUREG/CR-1756, 1982.

[10] R. E. Nightingale, Nuclear Graphite, Academic Press, New York, 1962.

[11] A. B. Chilton, J. K. Shultis, and R. E. Faw, Principles of Radiation Shielding, Prentice-Hall, Inc., 1984.

[12] R.L. Newacheck, H. Montgomery, R. E. Quilici, I. E. Lamb, and F. W. Boone, Aerojet-General Nucleonics Industrial Reactor: Hazards Summary Report, AN-1193, September 1964.

[13] Aerotest Operations, Inc., Updated Safety Analysis Report (USAR), Rev. 0, Docket 50-228, License No. R-98, undated.

[14] R. L. Tomlinson, Aerojet-General Nucleonics Industrial Reactor, AN-1405, April 1965.

[15] L. F. Mondolfo, Aluminum Alloys: Structure and Properties, Butterworths, London, 1976.

[16] National Institute of Standards and Technology, Compositions of Materials Library, http://physics.nist.gov/cgi-bin/Star/compos.pl?ap.

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[17] http://www.reade.com/products/35-oxides-metallic-powders/672-samariumiii-oxide-sm2o3-samarium-sesquioxide-samarium-oxide-powder-cas-12060-58 [18] Dimensions and static properties of aluminum I-beams-http://www.engineeringtoolbox.com/aluminum-i-beams-d 1326.html Page 45 of 47

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10. Spreadsheets Page 46 of 47

Rev. 0.0 The following spreadsheets are supplied with this report. The activities output by the ORIGEN code are in these spreadsheets. If the activities for other dates are desired, the activities computed for other decay times are provided in the spreadsheets. The discharge column refers to October 1, 2011 and the 1 year, 5 year and 10 year decay times refer to October 1, 2012, October 1, 2015, and October 1, 2020, respectively, assuming little to no operation after October 1, 2011.

Spreadsheet Contents Core shroud, top grid plate, bottom grid plate core support structure and table 19.xlsx and core shroud support stand Tbeams and Table 20.xlsx T-Beams Activities bolts and Table 21.xlsx Representative SS304 bolts in T-Beams Instrument Tubes and Table 22.xlsx Instrument Guide Tubes thermal column NRA graphite and Table Graphite in the thermal column and the NRA 23.xlsx Aluminum in representative locations of the Al_tra_tc activities and Table 24.xlsx thermal column and the NRA.

lead_activations and Table 25.xlsx Lead at the front and back of the NRA boral_activities and Table 26.xlsx Boral shutter activities Other AL_tankand Table 27.xlsx Aluminum tank activities Concrete below and nearest the core concrete_activities and Table 28.xlsx (bioshield) rebar_activities and table 29.xlsx Rebar in the concrete locations.

Soil under the core and behind the nearest side soils_activities and Table 30.xlsx wall.

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