ML18331A323

Calculation CALC-3016518, Revision 2, Radiation Effects on Materials for TMI-2 ISFSI License Renewal

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Issue date:	08/23/2018
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NA NA NA Orano Federal Services August 31, 2018 Records Management CALC-3016518-002 Page 1 of 43 0

Orano Federal Services CALCULATION orano Document No.:

02029.00.0000.02-01 Rev. No.

2 I

Page 1 of 43 Project No.:

02029.00.0000.02 Project Name: I NRC License Renewal

Title:

Radiation Effects on Materials for TMl-2 ISFSI License Renewal Summary:

The Independent Spent Fuel Storage Installation (ISFSI) for Three Mile Island Unit 2 (TMl-2) fuel was originally licensed for 20 years. The original 20 year license will expire in 2019, and the license will be extended an additional 20 years. The purpose of this calculation is to determine the effects on materials due to an additional 20 years of radiation exposure. Materials include steel of the TMl-2 Fuel, Knockout, and Filter Canisters, steel of the dry shielded canisters (DSCs), Licon within the TMl-2 Fuel Canisters, and concrete of the Horizontal Storage Modules (HSMs). It is concluded that there are no adverse effects on these materials due to the additional 20 years of storage.

This document is safety related.

Safety [gl Non-Safety D

Contains Unverified Input/ Assumptions:

Software Utilized:

MCNP SCALE Excel*

Software Active in FS EASI: Yes: [gl NA*: D

• Not Applicable per Section 5.7 of FS-EN-PRC-002 Error Reports & Associated Corrective Actions Reviewed: Yes: [gl No:D Printed Name Preparer:

SN Gibboney Checker:

EA Gonsiorowski Approver:

OS Hillstrom Other:

FS-EN-FRM-002 Rev. 10 (Effective March 1, 2018)

Reference:

FS-EN-PRC-002 Documentum Number: CALC-3016518-002 Yes:D No: [gj Version:

Storage Media: Yes: [gl No:D 5.1.51 and 6.1.00 6.0 2010 Location:

ColdStor Signature Date

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Page 2 of 43 Orano Federal Services

Title:

Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal REVISION HISTORY REV.

CHANGES 0

Original Issue 1

Correct MCNP stainless steel composition error.

Consider radiation effects on the Licon and stainless steel shrouds within the TMI-2 Fuel Canister.

2 Update design basis gamma source. This calculation revision supplies the technical basis to support the response to RAIs 3-2, 3-3, and 3-11 [36].

CALC-3016518-002 Page 2 of 43

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Title:

Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal TABLE OF CONTENTS TABLE OF CONTENTS......................................................................................................................... 3 LIST OF FIGURES.................................................................................................................................. 4 LIST OF TABLES.................................................................................................................................... 4 1.0 PURPOSE........................................................................................................................................ 5 2.0 METHODOLOGY......................................................................................................................... 5 3.0 ASSUMPTIONS.............................................................................................................................. 9 3.1 UNVERIFIED ASSUMPTIONS............................................................................................................ 9 3.2 JUSTIFIED ASSUMPTIONS................................................................................................................ 9 4.0 DESIGN INPUTS.......................................................................................................................... 10 5.0 SOURCE SPECIFICATION....................................................................................................... 11 6.0 CALCULATIONS........................................................................................................................ 15 6.1 MATERIAL PROPERTIES................................................................................................................ 15 6.2 ANALYSIS USING DESIGN BASIS SOURCES................................................................................... 20 6.3 ANALYSIS USING AMBECM SOURCE............................................................................................ 27 7.0 RESULTS/CONCLUSIONS........................................................................................................ 30 8.0 APPENDIX A................................................................................................................................ 32

8.1 REFERENCES

................................................................................................................................. 32 8.2 COMPUTER SOFTWARE USAGE..................................................................................................... 35 8.2.1 Computer Software Usage for Rev. 0................................................................................... 35 8.2.2 Computer Software Usage for Rev. 1................................................................................... 36 8.2.3 Computer Software Usage for Rev. 2................................................................................... 38 8.2.4 Excel 2010 File Listing......................................................................................................... 40 9.0 APPENDIX B................................................................................................................................ 41 CALC-3016518-002 Page 3 of 43

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal LIST OF FIGURES Figure 2.0 NUHOMS-12T Dry Shielded Canister............................................................................. 7 Figure 2.0 NUHOMS-12T Horizontal Storage Module..................................................................... 8 Figure 6.2 Design Basis Source Model, No Internals........................................................................ 25 Figure 6.2 Design Basis Source Model, With Internals..................................................................... 26 Figure 6.3 AmBeCm Source Model................................................................................................... 29 LIST OF TABLES Table 5.0 Design Basis Gamma Source Term (1998)........................................................................ 14 Table 5.0 Design Basis Gamma Source Term (1985)........................................................................ 15 Table 6.1 Fuel Mixture Composition................................................................................................. 17 Table 6.1 Stainless Steel Composition............................................................................................... 17 Table 6.1 Concrete Composition........................................................................................................ 17 Table 6.1 CA-25C Cement Composition........................................................................................... 18 Table 6.1 Borosilicate Glass (Pyrex) Composition............................................................................ 18 Table 6.1 Licon Composition, No Water........................................................................................... 18 Table 6.1 Licon Composition, Half of Nominal Water...................................................................... 19 Table 6.1 Licon Composition, Nominal Water.................................................................................. 19 Table 6.1 Licon Composition, Double Nominal Water...................................................................... 20 Table 6.2 Volume and Mass Computations for Tally Regions.......................................................... 23 Table 6.2 Results without TMI-2 Canister Internals.......................................................................... 23 Table 6.2 Results with TMI-2 Fuel Canister Internals....................................................................... 24 Table 6.3 AmBeCm Neutron Results................................................................................................. 28 Table 7.0 Gamma Results, Design Basis Source................................................................................ 31 Table 7.0 Neutron Results.................................................................................................................. 31 Table 9.0 Design Basis Gamma Source Term (2019)........................................................................ 42 CALC-3016518-002 Page 4 of 43

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal 1.0 Purpose The Independent Spent Fuel Storage Installation (ISFSI) for Three Mile Island Unit 2 (TMI-2) fuel was originally licensed for 20 years. The original 20 year license will expire in 2019, and the license will be extended an additional 20 years. The purpose of this calculation is to determine the effects on materials due to an additional 20 years of radiation exposure. Materials include steel of the TMI-2 Fuel, Knockout, and Filter Canisters, steel of the dry shielded canisters (DSCs), Licon within the TMI-2 Fuel Canisters, and concrete of the Horizontal Storage Modules (HSMs).

2.0 Methodology Twelve TMI-2 Canisters are stored in the DSC, and each DSC is stored inside an HSM, see Figure 2.0-1 and Figure 2.0-2, as extracted from Figure 1.2-4 and Figure 1.2-2 from the FSAR [7]. TMI-2 core debris is contained within three types of stainless steel TMI-2 Canisters: Fuel, Knockout, and Filter

[11,23,24]. While the internals of these three TMI-2 Canister types are different, the external radial dimensions of these three TMI-2 Canisters are the same. As a result, the first design basis source analysis did not model any internal hardware, producing an analysis applicable to all three types of TMI-2 Canisters. In the second design basis source analysis, some internals of the TMI-2 Fuel Canister are modeled in order to determine the radiation dosage absorbed by the stainless steel shroud for the BORAL and Licon, which is not applicable to the TMI-2 Knockout and Filter Canisters.

Radial models of the TMI-2 Fuel Canisters within an HSM are developed using MCNP [1] and design basis neutron and gamma source terms. The ends are neglected for simplicity because the maximum neutron flux and radiation exposure is typically at the sides rather than the ends. Also, the DSC features non-structural carbon steel shield plugs at each end that would highly attenuate gamma radiation travelling in the end directions and therefore reduce gamma energy deposition in the end concrete compared to the side concrete.

Two basic radial models are developed. In the first radial model, all internals of the TMI-2 Canister are conservatively ignored. Radiation damage is considered for gammas/neutrons in HSM concrete and neutrons in the TMI-2 Canister stainless steel and carbon steel of the DSC. In the second radial model, the internals of the TMI-2 Fuel Canister are evaluated. Radiation damage is considered for gammas/neutrons in the low-density concrete (Licon) within the TMI-2 Fuel Canister, as well as the stainless steel shrouds that support the BORAL poison plates. Licon is present only in the TMI-2 Fuel Canister. Computed radiation damage to the TMI-2 Fuel Canister stainless steel shrouds for the BORAL is indicative of the radiation damage accumulated in the stainless steel internals inside the TMI-2 Knockout and Filter Canisters because the TMI-2 Fuel Canister contains the general canister source terms.

Explicit radiation damage to the BORAL within the TMI-2 Fuel Canisters and B4C poison within the TMI-2 Knockout and Filter Canisters is not explicitly considered. Section 3.4.2.7 of [25] indicates, radiation embrittlement of borated aluminum alloys, aluminum metal-matrix composites, and BORAL is not expected to be credible. Also, the NRC has granted an exemption for providing positive means to verify the continued efficacy of neutron absorbing material. The justification is provided on page 4 of

[31], The neutron absorbing material (poison) is in a form that exposure to the ambient atmosphere of CALC-3016518-002 Page 5 of 43

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal the DSC interior will not cause a significant deterioration of the structural properties of the material over the expected life of the facility.

Spherical models of an Americium-Beryllium-Curium (AmBeCm) startup neutron source are also developed. According to Table 3.1-3 of the FSAR [7], two of the TMI-2 Fuel Canisters contain startup sources (see design input #8), which would create a highly localized neutron flux near the source.

Because the source could be anywhere inside a TMI-2 Fuel Canister, simple spherical models are developed to capture the peak neutron fluxes near the source.

The limit for absorbed dose for gammas in concrete is 1010 rad [3], and the limit for neutron fluence in concrete is 2x1019 n/cm2 [2], above which measurable degradation of concrete strength properties may occur. It is assumed that the radiation limits for standard concrete apply to Licon. The limit for neutron fluence in carbon steel is 1018 n/cm2 [Section 4.3.2.3 of [4],[22)) for alteration of steel mechanical properties. Neutron fluence limits are typically based on fast neutrons (E > 1.0 MeV) [22], although in this calculation the total neutron energy spectrum (E > 0 MeV) is conservatively computed. It is assumed that the limit for stainless steel is the same as carbon steel. No limit on gamma radiation damage applicable to steel has been identified, which is consistent with the Standardized NUHOMS System renewal application [5]. However, a limit of 3.8x1011 rad is imposed for possibly adversely affecting material properties of borated aluminum (BORAL) for gamma radiation exposure (Section 3.4.2.7 of [25]). Note that other limits may be found in literature for these materials, although the limits used are consistent with the limits in the Standardized NUHOMS System renewal application used for evaluating radiation effects of materials [5].

For gamma radiation, absorbed energy is computed by MCNP in units of MeV/(g-s) and converted to rad/s by multiplying by the conversion factor 1 MeV/g = 1.602x10-8 rad. The total absorbed energy is then obtained by multiplying by the total time (in seconds). For neutron radiation, flux is computed in units of n/(cm2-s) and converted to fluence by multiplying by the total time (in seconds).

The total irradiation time of the DSC steel and HSM concrete is 40 years (i.e., 20 years for the original license plus 20 years for the license extension). However, the TMI-2 Canisters were loaded as early as 1985 [6], while the first HSMs were loaded in 1999 [32], a gap of approximately 14 years. Therefore, the steel in the TMI-2 Canisters have experienced an additional 14 years (4.4182x108 s) of more intense irradiation compared to the DSC or HSM, for a total of 14+40 = 54 years. To be conservative, a gamma irradiation time of 60 years (1.8935x109 s) is assumed for all components using a 1998 decayed source term. In addition, 14 years of irradiation caused by exposure to a stronger gamma source (1985 decayed source term) is calculated separately, and then added to the 60-year cumulative dose for a total 74-year exposure period. This is an added level of conservatism for the gamma irradiation timeframe. Whereas, the neutron irradiation time of 60 years includes the initial 14-year TMI-2 Canister loading period and is used in the evaluation. This is justified for both 1998 bounding neutron sources (see Section 5.0), since the neutron source remains relatively constant with time from 1985 to 1998.

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal Figure 2.0 NUHOMS-12T Dry Shielded Canister CALC-3016518-002 Page 7 of 43

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal Figure 2.0 NUHOMS-12T Horizontal Storage Module CALC-3016518-002 Page 8 of 43

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal 3.0 Assumptions 3.1 Unverified Assumptions No unverified assumptions are utilized in the calculation.

3.2 Justified Assumptions Following is a list of justified assumptions:

1. The calculation is performed only in the radial direction because the maximum neutron flux and gamma exposure is typically larger at the sides compared to the ends because the source is long.

Also, the DSC features non-structural carbon steel shield plugs at each end that would highly attenuate gamma radiation travelling in the end directions and therefore reduce gamma energy deposition in the end concrete compared to the side concrete. Therefore, the ends of the TMI-2 Canisters, DSC, and HSM are ignored.

2. In all neutron calculations, the energy spectrum of the source is approximated with the spontaneous fission spectrum of Pu-240. Because the objective is simply to estimate the total neutron flux, the choice of spectrum has little effect on the results. Also, this approach is consistent with previous work [13,14]. In the MCNP models, the spontaneous fission spectrum of Pu-240 is represented by a continuous energy function per the MCNP manual [1].

3. It is assumed that the neutron fluence limit for carbon steel applies to stainless steel.

4. For the spent fuel in the TMI-2 Fuel Canisters, the source is assumed to be homogenously distributed throughout the available volume. This is a reasonable assumption for a collection of fuel debris. While the metallic structural components are included in the homogenized material description, because the fuel is debris, no attempt is made to divide the fuel into bottom nozzle, active fuel, plenum, and top nozzle regions, as the fuel debris is assumed to be mixed.

5. The HSM is modeled only in an approximate manner (either as a cylinder or sphere), although the actual geometry is a flat slab. This is acceptable because the neutron flux or gamma energy deposition is maximized on the inner surface of the HSM, and the models are conservatively developed so that the distance from the DSC to the HSM is consistent with the minimum distance of the actual geometry.

6. The actual size of the AmBeCm source is unknown but is assumed to be small. A radius of 1 cm is assumed based on an approximate source mass of 1 g Am and 6 g Be [13]. This is very small in comparison with the volume of a TMI-2 Fuel Canister. The MCNP model of the AmBeCm source is treated as concentric spheres of steel and concrete to capture the localized maxima from a concentrated source.

7. The total irradiation time of the DSC steel and HSM concrete is 40 years (i.e., 20 years for the original license plus 20 years for the license extension). However, the TMI-2 Canisters were loaded as early as 1985 [6], while the first HSMs were loaded in 1999 [32], a gap of approximately 14 years. Therefore, the steel in the TMI-2 Canisters have experienced an additional 14 years (4.4182x108 s) of more intense irradiation compared to the DSC or HSM, or a total of 14+40 = 54 years. To be conservative, a gamma irradiation time of 60 years (1.8935x109 s) is assumed for all components using a 1998 decayed source term. In addition, 14 years of irradiation caused by exposure to a stronger gamma source (1985 decayed source term) is calculated separately, and then added to the 60-year cumulative dose for a total 74-year exposure period. This is an added level of conservatism for the gamma irradiation timeframe.

CALC-3016518-002 Page 9 of 43

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal Whereas, the neutron irradiation time of 60 years includes the initial 14-year TMI-2 Canister loading period and is used in the evaluation. This is justified for both 1998 bounding neutron sources since the neutron source remains relatively constant with time from 1985 to 1998 (see Section 5.0).

8. Explicit radiation damage to the BORAL within the TMI-2 Fuel Canisters and B4C poison within the TMI-2 Knockout and Filter Canisters is not explicitly considered. Section 3.4.2.7 of

[25] indicates that radiation embrittlement of borated aluminum alloys, aluminum metal-matrix composites, and BORAL is not expected to be credible. Also, the NRC has granted an exemption for providing positive means to verify the continued efficacy of neutron absorbing material. The justification is provided on page 4 of [31], The neutron absorbing material (poison) is in a form that exposure to the ambient atmosphere of the DSC interior will not cause a significant deterioration of the structural properties of the material over the expected life of the facility.

9. It is assumed that the radiation limits for standard concrete apply to Licon.

10. Because both the TMI-2 Fuel Canisters and DSC are vented to the atmosphere, over time moisture may re-enter the TMI-2 Fuel Canisters [29]. It is assumed this moisture could re-hydrate the Licon.

11. Licon is nominally 11 wt.% soda-lime-borosilicate glass bubbles [7,27]. The composition of the soda-lime-borosilicate glass bubbles is not precisely known, although a reasonable approximation is borosilicate glass (Pyrex) [30].

4.0 Design Inputs Following is a list of major design input documents. Minor design inputs are addressed as-needed in the calculation.

1. The design basis gamma source term is obtained from Table 7.2-2 of the Final Safety Analysis Report (FSAR) [7] and is reproduced in Table 5.0-1. This source term represents decay to March 1998. Also, an additional gamma source term is obtained from source.out and shown in Table 5.0-2. This source term represents decay to 1985 and will be applied to only the TMI-2 Canisters (not the DSC or HSM), which were already loaded with debris from 1985 to 1999, at which point final loading into the DSC and HSM occurred. Note that the FSAR source term was calculated in March 1998, prior to the actual loading of the DSCs and HSMs in 1999. The difference in source strength between the source term in the FSAR and the actual load date is determined to have a minimal impact on results.

2. The design basis neutron source term is obtained from Table 7.2-1 of the FSAR [7] and is for each TMI-2 Canister 6.895x105 n/s. This source term represents decay to March 1998.

However, footnote (1) to Table 7.2-1 indicates that two of the TMI-2 Fuel Canisters may contain an AmBeCm startup source with a magnitude per TMI-2 Canister of 7.3x106 n/s, or a total neutron source per TMI-2 Canister of 8x106 n/s. This is a factor of 8x106/6.895x105 ~ 12 times larger than the design basis neutron source.

3. The DSC basket geometry is defined on the applicable FSAR drawing [8].

4. The DSC shell geometry is defined on the applicable FSAR drawing [9].

5. The HSM geometry is defined on the applicable FSAR drawing [10].

6. The TMI-2 Fuel Canister geometry is defined on the applicable FSAR drawing [11].

CALC-3016518-002 Page 10 of 43

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal

7. An ORIGEN-ARP input file is developed to determine how the design basis sources change with time. The B&W 15x15 library is used. Decay times from 0 to 60 years are requested. Per the original source term calculation [21], the peak assembly burnup is 3,175 MWd/MTU x 1.879 =

5,966 MWd/MTU, and the cycle length is 117 days. The fuel loading is 0.472 MTU per fuel assembly [21]. Because the TMI-2 Fuel Canister may contain more than a single fuel assembly, the fuel loading is scaled by 1.259 [21], or 0.472*1.259 = 0.594 MTU. The enrichment is conservatively selected to be 1.98 wt.% U-235 [21]. Therefore, the assembly power to produce the desired burnup is 5,966/117*0.594 = 30.29 MW.

8. Two startup neutron sources are present in the TMI-2 Canisters. According to Table 3.1-3 of the FSAR [7] and page 9 of NRC inspection report 72-20/01-02 [32], the startup sources are located in DSC 1/TMI-2 Canister D-196 and DSC 5/TMI-2 Canister D-122. Because D-196 and D-122 are both TMI-2 Fuel Canisters, the startup neutron source fluence analysis is performed only for the TMI-2 Fuel Canister.

5.0 Source Specification Both gamma and neutron source terms are calculated and decayed from the time of the Three Mile Island accident (March 1979) to March 1985, or 6 years. This decay time represents the passage of time until the approximate initial loading date of the TMI-2 Canisters [6]. These 6-year source terms are then used in the MCNP models to determine the energy deposition by the gamma and neutron sources into the TMI-2 Canisters. A second set of source terms decayed from 1979 to 1998 is used when the TMI-2 Canisters were loaded into the DSCs and HSMs approximately 14 years after the initial loading of the TMI-2 Canisters [32], which are obtained from Table 7.2-1 and 7.2-2 of the FSAR [7]. The total irradiation time of the DSC steel and HSM concrete is 40 years (i.e. 20 years for the original license plus 20 years for the license extension), but the total irradiation time of the TMI-2 Canisters include the 14 years prior to loading into the DCS and HSM. For the gamma source, an additional 14 years of gamma irradiation caused by exposure to an initially stronger source (1985 decayed source term) is calculated and then conservatively added to the cumulative gamma dose accrued over the entire 60 years for a total 74-year exposure period. This is an added level of conservatism for the gamma irradiation timeframe.

Whereas, the neutron irradiation time of 60 years includes the initial 14-year TMI-2 Canister loading period and is used in the evaluation. This is justified for both 1998 bounding neutron sources since the neutron source remains relatively constant with time from 1985 to 1998.

Design Basis Gamma Source The design basis gamma source in 1998 is provided on a per-TMI-2 Canister basis and is reproduced in Table 5.0-1. Because there are up to 12 TMI-2 Canisters inside each DSC, the total gamma source is 6.372x1014 x 12 = 7.646x1015 /s per DSC. This gamma source applied to the DSC and HSM conservatively estimates the energy absorbed by the concrete over a 60 year time period because no credit is taken for the reduction of the gamma source strength from 1998 due to decay over this time period.

The design basis gamma source in 1985 is provided on a per-TMI-2 Canister basis and is in Table 5.0-2.

Only the TMI-2 Canisters have the gamma source strength in 1985 applied from 1985 until 1999, with the total gamma source being 8.226x1014 /s per TMI-2 Canister over this period. This initial gamma dose accumulation is calculated prior to loading into the DSCs, and then added to the calculated gamma CALC-3016518-002 Page 11 of 43

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal dose accumulation over 60 years after loading into the DSCs and HSMs (74-year total exposure period).

The gamma source is distributed throughout the interior of each TMI-2 Canister.

Design Basis Neutron Source Two neutron source terms are considered. The design basis neutron source is 6.895x105 n/s per TMI-2 Canister, or 6.895x105 x 12 = 8.274x106 n/s per DSC. As the energy spectrum of Table 7.2-1 of [7] is representative of spontaneous fission of Pu-240, the MCNP continuous energy function for Pu-240 fission is used. Note that estimation of total flux is not highly dependent on the spectrum input. The neutron source is distributed throughout the interior of each TMI-2 Canister.

The neutron source is representative of the source in 1998, although the TMI-2 Canisters were loaded as early as 1985. Therefore, the change of the neutron source with time is investigated to determine if it is appropriate to use the design basis neutron source over the time period from 1985 to 1998 for irradiation of the TMI-2 Canisters. The neutron source was originally computed using the ORIGEN2 program, which is no longer used. Therefore, an ORIGEN-ARP [17] input file is developed based on information provided in the FSAR [7] and the original source term calculation [21]. Per the original source term calculation [21], the peak assembly burnup is 3,175 MWd/MTU x 1.879 = 5,966 MWd/MTU, and the cycle length is 117 days. The fuel loading is 0.472 MTU per fuel assembly [21]. Because the TMI-2 Fuel Canister may contain more than a single fuel assembly, the fuel loading is scaled by 1.259 [21], or 0.472*1.259 = 0.594 MTU. The enrichment is conservatively selected to be 1.98 wt.% U-235 [21].

Therefore, the assembly power to produce the desired burnup is 5,966/117*0.594 = 30.29 MW. An ORIGEN-ARP input file is developed with these input parameters and a B&W 15x15 library. Decay times from 0 to 60 years are requested. Note that the TMI-2 Fuel Canister contains the general source term calculated without canister internals.

The results are provided in file source.out. Due to the very low burnup, the change in the neutron source with time is quite different than typical spent fuel. The neutron source in 1985 (6.618x105 n/s at 6 years decay) is actually smaller than the design basis neutron source in 1998 (6.965x105 n/s at 19 years decay). The reason is that as Pu-241 (a beta emitter) decays to Am-241 (an alpha emitter), the (,n) neutron source component grows, while the spontaneous fission source component decreases. The net effect is that the neutron source remains relatively constant with time over the period of interest.

Therefore, it is acceptable to use the design basis neutron source for the time period from 1985 to 1998.

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal AmBeCm Neutron Source The neutron source term for the two TMI-2 Fuel Canisters that contain AmBeCm startup source material is 8x106 n/s per TMI-2 Fuel Canister (6.895x105 n/s for fuel + 7.3x106 n/s for AmBeCm). While this source bounds the design basis neutron source by over a factor of 10, it affects only two of the 341 loaded TMI-2 Canisters per the licensing evaluation for the AmBeCm source material [12]1. Therefore, this source, while large, is not representative of the typical spent fuel sources and is treated separately.

For simplicity, the Pu-240 spontaneous fission spectrum is applied to the AmBeCm source, as the estimation of total flux is not highly dependent on the spectrum input. Also, [12] indicates the Pu-240 spontaneous fission spectrum was used when estimating dose rates from the AmBeCm source.

The AmBeCm source strength is estimated to be 7.3x106 n/s per footnote (1) of Table 7.2-1 of the FSAR

[7]. This source strength represents decay of approximately 23 years from the time of the accident per the source derivation in EDF-1793 [13]. Since the accident occurred in 1979, and the earliest TMI-2 Fuel Canisters were loaded in 1985 (6 years decay), a review of the AmBeCm source provided in the FSAR for this calculation is undertaken given that the TMI-2 Fuel Canisters are irradiated for a longer time period than both the DSC and HSM. At the time of the accident in 1979, it is conservatively estimated that the AmBeCm source contained 3.4 Ci Am-241 and 450 Ci Cm-242 [13]. Am-241 has a half-life of 432.7 years, while Cm-242 has a half-life of 0.446 years [19]. After 6 years of decay, the source would contain 3.4 Ci Am-241 (essentially unchanged) but only 0.04 Ci Cm-242. Because the Cm-242 activity at the time the TMI-2 Fuel Canisters were loaded is quite small, the neutron source strength at 6 years decay is similar to the neutron source strength reported in the FSAR. Therefore, it is acceptable to use the 7.3x106 n/s source strength as a representative value for the AmBeCm source at the time of TMI-2 Fuel Canister loading. In the MCNP model, a total source of 8x106 n/s is used, which includes the spent fuel source.

1 While [12] indicates 344 total TMI-2 Canisters, the total number stored at the TMI-2 ISFSI is 341, per the license [20]. The remaining three TMI-2 Canisters were never stored at the TMI-2 ISFSI.

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal Table 5.0 Design Basis Gamma Source Term (1998)

Eupper (MeV)

Source per TMI-2 Canister

(/s) 5.00E-02 2.679E+14 1.00E-01 4.836E+13 2.00E-01 2.068E+13 3.00E-01 9.887E+12 4.00E-01 5.260E+12 6.00E-01 1.773E+14 8.00E-01 9.471E+13 1.00E+00 1.364E+12 1.33E+00 8.250E+12 1.66E+00 3.504E+12 2.00E+00 3.533E+10 2.50E+00 7.066E+07 3.00E+00 7.196E+06 4.00E+00 1.924E+05 5.00E+00 1.027E+04 6.50E+00 8.988E+03 8.00E+00 1.523E+03 1.00E+01 2.401E+02 Totals 6.372E+14 CALC-3016518-002 Page 14 of 43

Page 15 of 43 Orano Federal Services

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal Table 5.0 Design Basis Gamma Source Term (1985)

Eupper (MeV)

Source per TMI-2 Canister

(/s) 5.00E-02 2.605E+14 1.00E-01 7.674E+13 2.00E-01 5.865E+13 3.00E-01 1.678E+13 4.00E-01 1.232E+13 6.00E-01 4.149E+13 8.00E-01 3.365E+14 1.00E+00 1.282E+13 1.33E+00 4.826E+12 1.66E+00 1.349E+12 2.00E+00 1.569E+11 2.50E+00 4.668E+11 3.00E+00 9.521E+09 4.00E+00 8.608E+08 5.00E+00 1.935E+04 6.50E+00 7.711E+03 8.00E+00 1.503E+03 1.00E+01 3.177E+02 Totals 8.226E+14 6.0 Calculations 6.1 Material Properties Material properties for the fuel, stainless steel, and concrete used in the MCNP models are obtained from Tables 4, 5, and 6 of the original Vectra shielding calculation [14].

The fuel debris is a mixture of UO2, zircaloy, stainless steel, and Inconel per Table 3 of [14]. The density (g/cm3) of each element in the fuel mixture from Table 4 of [14] is used to compute the weight percent of each element in the fuel mixture. This composition is provided in Table 6.1-1. However, the mixture density cannot be directly used from Table 4 of [14] because in that reference, the total mass from 12 TMI-2 Canisters is homogenized over a large volume that includes the space between TMI-2 Canisters, resulting in a low fuel mixture density. Therefore, the mixture density is computed based on the TMI-2 Canister geometry used in the current MCNP models.

Per Table 3 of [14], the mass of fuel debris within each TMI-2 Fuel Canister is 862,405 g. In the current calculation, the TMI-2 Fuel Canister is modeled with a length of 135-in, corresponding to the BORAL CALC-3016518-002 Page 15 of 43

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal length on drawing [11]. The inner diameter of the TMI-2 Fuel Canister 14 - 0.25*2 = 13.5-in [11]. The internal volume of a TMI-2 Fuel Canister in the MCNP models when TMI-2 Fuel Canister internals are ignored is then 316,659 cm3, or a fuel density of 862,405/316,659 = 2.72 g/cm3.

For the cases in which the TMI-2 Fuel Canister internals are explicitly modeled, the fuel mixture is constrained to the region inside the inner shroud, which has an inner square dimension of 9-in [11]. The length is modeled as 135-in, consistent with the models without TMI-2 Fuel Canister internals. The internal fuel volume is then 179,193 cm3, or a fuel density of 862,405/179,193 = 4.81 g/cm3. The fuel density is larger to preserve the fuel mass, as the available volume is smaller when TMI-2 Fuel Canister internals are included in the models.

The stainless steel composition and density are obtained from Table 5 of [14] and are reproduced in Table 6.1-2. It is used for both the stainless steel TMI-2 Fuel Canisters and carbon steel DSC. The effect of modeling the DSC as stainless steel rather than carbon steel is negligible, particularly since both materials are modeled with a density of 7.85 g/cm3 in the design basis dose rate calculations [14].

The concrete composition is obtained from Table 6 of [14] and is reproduced in Table 6.1-3. Per [14],

the concrete density is 145 lb/ft3, which is converted to 2.322 g/cm3 as 145/2.205*1000/123/2.543.

Low-density concrete, or Licon, is used between the outer stainless steel shroud and the inner diameter of the TMI-2 Fuel Canister in the models in which the internals are included [11]. Licon is nominally 60 wt.% CA-25C refractory cement, 11 wt.% soda-lime-borosilicate glass bubbles, and 29 wt. % water

[7,27]. The dried density is nominally 52 lb/ft3 [27], or 0.833 g/cm3, which is used as the nominal density in the MCNP models. The green and soaked densities of 64 lb/ft3 and 62.5 lb/ft3 from [27] are not directly used in the MCNP models, although a sensitivity analysis on water content is performed that encompasses other Licon densities because the water composition of the Licon within the TMI-2 Fuel Canisters is not precisely known. When the TMI-2 Fuel Canisters are vacuum dried, some of the water will be driven out of the Licon [28]. Also, because both the TMI-2 Fuel Canisters and DSC are vented to the atmosphere, over time moisture may re-enter the TMI-2 Fuel Canisters [29]. It is assumed this moisture could re-hydrate the Licon. Because the precise water content of the Licon is not known and could vary between TMI-2 Fuel Canisters, four different water contents are considered: (1) no water, (2) half of the nominal water content, (3) nominal water content, and (4) twice the nominal water content.

The typical composition of CA-25C cement is summarized in Table 6.1-4 [27]. Note that 0.5 wt.% is added to the Al2O3 component provided in [27] so that the total sums to 100 wt.%. The composition of the soda-lime-borosilicate glass bubbles is not precisely known, although a reasonable approximation is borosilicate glass (Pyrex). The composition of borosilicate glass is provided in Table 6.1-5 [30]. Based on the stated nominal density and constituent weight percents, the density of each component may be computed for the four scenarios considered. The full details are provided in spreadsheet Data_R1.xlsx, sheet Licon, and the Licon composition for the four water contents considered are provided in Table 6.1-6 though Table 6.1-9. Note that the densities of the cement and glass in the mixture are treated as fixed quantities at their nominal values.

CALC-3016518-002 Page 16 of 43

Page 17 of 43 Orano Federal Services

Title:

Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal Table 6.1 Fuel Mixture Composition Element Weight Percent O

9.317 Cr 0.731 Fe 2.050 Ni 0.921 Zr 17.951 U-235 1.367 U-238 67.662 Density = 2.72 g/cm3 (models without internal hardware)

Density = 4.81 g/cm3 (models with internal hardware)

Table 6.1 Stainless Steel Composition Element Weight Percent Cr 19.0 Fe 71.75 Ni 9.25 Density = 7.85 g/cm3 Table 6.1 Concrete Composition Element Weight Percent H

0.55 O

49.8 Na 1.70 Al 4.55 Si 31.6 K

1.91 Ca 8.26 Fe 1.23 Density = 2.322 g/cm3 CALC-3016518-002 Page 17 of 43

Page 18 of 43 Orano Federal Services

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal Table 6.1 CA-25C Cement Composition Constituent Weight Percent Al2O3 80.2 CaO 18.4 MgO 0.4 SiO2 0.2 Fe2O3 0.3 Na2O 0.5 0.5% added to the reference value of 79.7% to result in a total of 100%.

Table 6.1 Borosilicate Glass (Pyrex) Composition Element Weight Percent B

3.7 Al 1.0 O

53.5 Si 37.7 Na 4.1 Table 6.1 Licon Composition, No Water Element Water (g/cm3)

Cement (g/cm3)

Glass (g/cm3)

Combined (g/cm3)

H 0.0000E+00 0.0000E+00 O

0.0000E+00 2.1731E-01 4.9011E-02 2.6632E-01 Al 2.1210E-01 9.1610E-04 2.1301E-01 Ca 6.5711E-02 6.5711E-02 Mg 1.2053E-03 1.2053E-03 Si 4.6715E-04 3.4537E-02 3.5004E-02 Fe 1.0485E-03 1.0485E-03 Na 1.8535E-03 3.7560E-03 5.6095E-03 B-10 6.2473E-04 6.2473E-04 B-11 2.7648E-03 2.7648E-03 Total 0.0000E+00 4.9969E-01 9.1610E-02 5.9130E-01 CALC-3016518-002 Page 18 of 43

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal Table 6.1 Licon Composition, Half of Nominal Water Element Water (g/cm3)

Cement (g/cm3)

Glass (g/cm3)

Combined (g/cm3)

H 1.3513E-02 1.3513E-02 O

1.0725E-01 2.1731E-01 4.9011E-02 3.7356E-01 Al 2.1210E-01 9.1610E-04 2.1301E-01 Ca 6.5711E-02 6.5711E-02 Mg 1.2053E-03 1.2053E-03 Si 4.6715E-04 3.4537E-02 3.5004E-02 Fe 1.0485E-03 1.0485E-03 Na 1.8535E-03 3.7560E-03 5.6095E-03 B-10 6.2473E-04 6.2473E-04 B-11 2.7648E-03 2.7648E-03 Total 1.2076E-01 4.9969E-01 9.1610E-02 7.1206E-01 Table 6.1 Licon Composition, Nominal Water Element Water (g/cm3)

Cement (g/cm3)

Glass (g/cm3)

Combined (g/cm3)

H 2.7025E-02 2.7025E-02 O

2.1449E-01 2.1731E-01 4.9011E-02 4.8081E-01 Al 2.1210E-01 9.1610E-04 2.1301E-01 Ca 6.5711E-02 6.5711E-02 Mg 1.2053E-03 1.2053E-03 Si 4.6715E-04 3.4537E-02 3.5004E-02 Fe 1.0485E-03 1.0485E-03 Na 1.8535E-03 3.7560E-03 5.6095E-03 B-10 6.2473E-04 6.2473E-04 B-11 2.7648E-03 2.7648E-03 Total 2.4152E-01 4.9969E-01 9.1610E-02 8.3282E-01 CALC-3016518-002 Page 19 of 43

Page 20 of 43 Orano Federal Services

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal Table 6.1 Licon Composition, Double Nominal Water Element Water (g/cm3)

Cement (g/cm3)

Glass (g/cm3)

Combined (g/cm3)

H 5.4051E-02 5.4051E-02 O

4.2898E-01 2.1731E-01 4.9011E-02 6.9530E-01 Al 2.1210E-01 9.1610E-04 2.1301E-01 Ca 6.5711E-02 6.5711E-02 Mg 1.2053E-03 1.2053E-03 Si 4.6715E-04 3.4537E-02 3.5004E-02 Fe 1.0485E-03 1.0485E-03 Na 1.8535E-03 3.7560E-03 5.6095E-03 B-10 6.2473E-04 6.2473E-04 B-11 2.7648E-03 2.7648E-03 Total 4.8303E-01 4.9969E-01 9.1610E-02 1.0743E+00 6.2 Analysis Using Design Basis Sources TMI-2 Canister without Internal Hardware In the first set of MCNP models, no hardware internal to the TMI-2 Canister is modeled. In these models, the fuel with design basis sources is conservatively in contact with the wall of the TMI-2 Canister, which increases the computed radiation damage to the TMI-2 Canister wall, the DSC, and the HSM. In the actual TMI-2 Fuel Canister, a layer of Licon is present between the fuel and TMI-2 Fuel Canister wall, which would absorb some of the energy. The gamma energy deposition in HSM concrete and the neutron fluence in steel/HSM concrete is computed using the MCNP5 v1.51 computer program

[1]. The gamma energy deposition in the TMI-2 Canisters from 1985 to 1999 is computed using the MCNP6.1 computer program [34]. All relevant dimensions are modeled, although only radial features are considered because gamma energy deposition and neutron fluence typically are a maximum in the radial direction. The model geometry is illustrated in Figure 6.2-1.

The axial length of the models is consistent with the approximate cavity length of the BORAL in the TMI-2 Fuel Canister cavity, 135-in [11]. Beyond this length, the ends are modeled as 1 ft. of concrete simply to add reflection. Each TMI-2 Fuel Canister has an outer diameter of 14-in and wall thickness of 0.25-in [11].

Each DSC contains up to 12 TMI-2 Canisters, 3 in an inner row and 9 in an outer row, as defined in the basket drawing [8]. The centerlines of the 3 inner TMI-2 Canisters are at a diameter of 18.75-in, while the centerlines of the 9 outer TMI-2 Canisters are at a diameter of 48.6-in.2 The inner diameter of the DSC is 65.94-in with a wall thickness of 0.63-in [9].

The HSM concrete is modeled only in an approximate manner because both energy deposition and neutron flux is maximized on the inner surface of the HSM, while the HSM itself is thick. The inner width of the HSM is 6 ft. 3-in [10] and the outer diameter of the DSC is 67.19-in [9]. Therefore, the 2 Modeled as 48.5-in, although this difference has a negligible effect on the result.

CALC-3016518-002 Page 20 of 43

Page 21 of 43 Orano Federal Services

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal minimum radial distance from the DSC to the HSM is (75 - 67.19)/2 = 3.905-in. The side of the HSM is modeled as an annulus of concrete 3.9-in from the DSC and is approximately 1 ft thick. The gamma energy deposition and neutron flux tallies are performed on an inner annulus of concrete 1 cm thick.

The source is modeled as distributed throughout the active length of the TMI-2 Canister. No credit is taken for the internal structure of the TMI-2 Canister. Therefore, the source is in direct contact with the TMI-2 Canister shell inner surface.

In the gamma input file, energy deposition is tallied on the inner 1 cm of the HSM concrete using an MCNP F6 tally. The default unit of an MCNP F6 tally is MeV/(g-s). 1 MeV = 1.602x10-6 erg [33], and 1 rad = 100 erg/g [33]. Therefore, 1 MeV/g = 1.602x10-8 rad. The tally is converted to rad per 60 years by multiplying by the number of seconds in a year, or 1.8935x109 s. The final conversion factor is then (1.602x10-8 rad/(MeV/g)) x (1.8935x109 s) = 30.333 rad-s-g/MeV, which is applied on the MCNP FM card. The computed energy deposition is 1.98x108 rad, which is below the limit of 1010 rad. This result is summarized in Table 6.2-2.

For completeness, gamma energy deposition is also computed in the TMI-2 Canister shell and DSC shell, although the energy deposition is not limited from a radiation damage standpoint. MCNP is not able to determine the mass of the TMI-2 Canister shell of 1.875x105 g, which is input on the MCNP SD card. This mass is computed based on a length of 135-in, an outer diameter of 14-in, a wall thickness of 0.25-in, and a density of 7.85 g/cm3. The volume and density calculations are also provided in spreadsheet Data_R1.xlsx, sheet Mass, and are summarized in Table 6.2-1. To calculate the initial gamma energy deposition in the TMI-2 Canister between 1985 and 1999, an MCNP F6 tally is used.

The tally is converted to rad per 14 years by multiplying by the number of seconds in a year, or 4.4182 x108 s. The final conversion factor is then (1.602x10-8 rad/(MeV/g)) x (4.4182x108 s) = 7.0780 rad-s-g/MeV, which is applied on the MCNP FM card. The initial results are then added to the results from the MCNP model using the 1998 source following the 60-year gamma energy deposition. Together, this produces the cumulative gamma dose in the TMI-2 Canisters. Therefore, the total gamma energy deposition in the TMI-2 Canister shell is 1.40x109 rad + 6.02x108 rad = 2.00x109 rad. The gamma energy deposition in the TMI-2 DSC shell is 4.42x108 rad. Note that the energy deposition in the inner circular array of three TMI-2 Canisters is approximately 20% larger than the outer array of nine TMI-2 Canisters. Therefore, the maximum energy deposition in the inner array of three is reported.

In the neutron input file, the flux is tallied on the shell of each TMI-2 Canister, on the DSC shell, and on the inner 1 cm of the HSM concrete using MCNP F4 tallies. As the default unit of an MCNP F4 tally is n/(cm2-s), the tally is converted to fluence by multiplying by 1.8935x109 s, which is applied on the MCNP FM card. MCNP is not able to determine the volume of the TMI-2 Canister shell of 23,891 cm3, which is input on the MCNP SD card. This volume is computed based on a length of 135-in, an outer diameter of 14-in, and a wall thickness of 0.25-in. The volume calculation is also provided in spreadsheet Data_R1.xlsx, sheet Mass, and is summarized in Table 6.2-1. The total neutron fluence (E

> 0 MeV) in the TMI-2 Canister and DSC is 9.23x1011 n/cm2 and 5.67x1011 n/cm2, respectively, which are below the fast fluence limit of 1018 n/cm2 (E > 1 MeV). These results are summarized in Table 6.2-2. Note that the fluence in the inner circular array of three TMI-2 Canisters is approximately 25%

larger than the outer array of nine TMI-2 Canisters. Therefore, the maximum fluence in the inner array of three is reported. The total neutron fluence (E > 0 MeV) in the concrete is 5.49x1011 n/cm2, which is below the fast fluence limit of 1019 n/cm2 (E > 1 MeV).

CALC-3016518-002 Page 21 of 43

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal TMI-2 Fuel Canister with Internal Hardware In the second set of MCNP models, the internals of the TMI-2 Fuel Canister are modeled to determine the radiation damage in the stainless steel shroud and Licon. The internals of the TMI-2 Fuel Canister consist of a 0.04-in thick inner stainless steel shroud, 0.135-in thick BORAL, and 0.08-in thick outer shroud [26]. The inner and outer stainless steel shroud are square, and BORAL plates are sandwiched between the shrouds. Licon fills the region between the outer shroud and the wall of the TMI-2 Fuel Canister.

When the internal TMI-2 Fuel Canister hardware is modeled, energy deposition/fluence in the TMI-2 Fuel Canister wall is computed simply to verify that it is bounded by the values computed when the internal TMI-2 Fuel Canister hardware is ignored. However, the energy deposition/fluence is not computed in the DSC or HSM because it is bounded by the results in Table 6.2-2.

For simplicity, only the inner shroud and Licon are explicitly modeled. The radiation damage of the inner shroud bounds the outer shroud because it is closer to the source material. The BORAL plates are not explicitly modeled because it is stated in Section 3.2, Justified Assumptions, that radiation embrittlement of BORAL plates is not a concern.

The width of the inner compartment is modeled as 9-in [11]. The inner shroud is modeled as 0.04-in thick (Page D.3 of [7], [26]). The model geometry is illustrated in Figure 6.2-2. Licon is modeled in close contact to the inner shroud, as the BORAL plates and outer shroud are not modeled for simplicity. Licon is modeled in two regions, an inner region 0.197-in thick, and an outer region that extends to the TMI-2 Fuel Canister wall. The inner region of the as-modeled Licon corresponds approximately to the location of the BORAL plates and outer shroud. Radiation damage in the Licon is conservatively estimated in the 0.197-in thick inner region, which is closer to the source material.

As noted in Table 6.1-1, the density of the fuel mixture is increased to preserve mass in the models that include the internal hardware because the available volume is smaller.

Because the water composition of the Licon is not known precisely, the models are run for four different water compositions: (1) no water, (2) half of the nominal water content, (3) nominal water content, and (4) twice the nominal water content. The range of Licon water compositions examined is sufficient to determine if the water content has a significant effect on the results. The Licon compositions are provided in Table 6.1-6 though Table 6.1-9.

In the gamma models, energy deposition is computed using MCNP F6 tallies, and in the neutron models flux is computed using MCNP F4 tallies. In the gamma tallies, the mass of each cell is a required input, which changes for the various Licon compositions. In the neutron tallies, the volume of each cell is a required input. The volume and density calculations are provided in spreadsheet Data_R1.xlsx, sheet Mass, and are summarized in Table 6.2-1.

Results are summarized in Table 6.2-3. Consistent with the results without internals, the TMI-2 Fuel Canisters located in the inner region of the DSC have larger energy deposition/fluence values than the TMI-2 Fuel Canisters in the outer region of the DSC. Only maximum values are reported in Table 6.2-3. It is observed that the mass of water in the Licon has little effect on the gamma energy deposition within the Licon. This is an expected result, as a rad is energy absorbed per unit mass. While total absorbed energy (MeV) will increase with increasing water density, the energy absorbed per unit mass (MeV/g) is relatively constant. Note that the energy absorbed in the TMI-2 Fuel Canister wall is less CALC-3016518-002 Page 22 of 43

Page 23 of 43 Orano Federal Services

Title:

Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal when the Licon is included (compare Table 6.2-2 to Table 6.2-3) because the Licon absorbs some of the energy. The absorbed energy in the Licon of 1.61x109 rad is less than the limit of 1010 rad.

For the neutron calculations, the fluence shows a clear increase as the water density in the Licon is reduced. Because water acts as a neutron shield, less water may result in more neutron interactions between TMI-2 Fuel Canisters. The total neutron fluence in the shroud and Licon is 8.98x1011 n/cm2 and 8.92x1011 n/cm2, which is less than the limits of 1018 n/cm2 and 2x1019 n/cm2, respectively. Note that the fluence in the TMI-2 Fuel Canister wall is less when the Licon is included (compare Table 6.2-2 to Table 6.2-3) because the Licon absorbs some of the neutrons.

Table 6.2 Volume and Mass Computations for Tally Regions Volume (cm3)

Density (g/cm3)

Mass (g)

Shroud 3.200E+03 7.85 2.512E+04 Licon No Water 1.616E+04 0.59130 9.555E+03 Licon 1/2 Water 1.616E+04 0.71206 1.151E+04 Licon Nominal Water 1.616E+04 0.83282 1.346E+04 Licon 2x Nominal Water 1.616E+04 1.07433 1.736E+04 TMI-2 Fuel Canister wall 2.389E+04 7.85 1.875E+05 DSC wall Computed by MCNP Automatically HSM wall Computed by MCNP Automatically Table 6.2 Results without TMI-2 Canister Internals Gamma Results (rad)

Region Value TMI-2 Canister steel shell 2.00E+09 DSC steel 4.42E+08 HSM concrete 1.98E+08 Neutron Results (n/cm2)

Region Value TMI-2 Canister steel shell 9.23E+11 DSC steel 5.67E+11 HSM concrete 5.49E+11 CALC-3016518-002 Page 23 of 43

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal Table 6.2 Results with TMI-2 Fuel Canister Internals Gamma Results (rad)

Region Licon No Water Licon 1/2 Nominal Water Licon Nominal Water Licon 2x Nominal Water Maximum Shroud steel 1.96E+09 1.96E+09 1.95E+09 1.93E+09 1.96E+09 Licon 1.61E+09 1.61E+09 1.61E+09 1.59E+09 1.61E+09 TMI-2 Fuel Canister steel shell 1.42E+09 1.38E+09 1.34E+09 1.27E+09 1.42E+09 DSC steel Not computed, bounded by Table 6.2-2 HSM concrete Not computed, bounded by Table 6.2-2 Neutron Results (n/cm2)

Region Licon No Water Licon 1/2 Nominal Water Licon Nominal Water Licon 2x Nominal Water Maximum Shroud steel 8.98E+11 7.31E+11 6.07E+11 4.53E+11 8.98E+11 Licon 8.92E+11 7.24E+11 6.00E+11 4.44E+11 8.92E+11 TMI-2 Fuel Canister steel shell 8.55E+11 6.82E+11 5.54E+11 3.90E+11 8.55E+11 DSC steel Not computed, bounded by Table 6.2-2 HSM concrete Not computed, bounded by Table 6.2-2 Computed only to verify that results are bounded by Table 6.2-2.

CALC-3016518-002 Page 24 of 43

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal x-y view x-z view Figure 6.2 Design Basis Source Model, No Internals CALC-3016518-002 Page 25 of 43

Page 26 of 43 Orano Federal Services

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal x-y view x-z view Figure 6.2 Design Basis Source Model, With Internals CALC-3016518-002 Page 26 of 43

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal 6.3 Analysis Using AmBeCm Source The spent fuel source within a TMI-2 Fuel Canister is volumetrically distributed. However, the precise size of the AmBeCm neutron source is unknown but is assumed to be small. Because the AmBeCm source is small, concentrated, and has a neutron source ~12x that of the design basis neutron source, it is not appropriate to smear the source throughout the volume of a TMI-2 Fuel Canister because the effect of such a source would be localized. Therefore, a pseudo-model is developed consisting of concentric spheres with an AmBeCm source in the center, see Figure 6.3-1. The pseudo-model is developed only for the TMI-2 Fuel Canister including the internals because the AmBeCm source is only in a TMI-2 Fuel Canister (see design input #8). Because the AmBeCm source is modeled in contact with the inner shroud of the TMI-2 Fuel Canister, the pseudo-model simulates the scenario in which the source is off-center, which maximizes the fluence in the surrounding structures.

The source contains approximately 1 g of Am-241 and 6 g of Be [13]. Given an Am density of 13.67 g/cm3 and a Be density of 1.848 g/cm3 [18], the approximate volume of source material is 1/13.67+6/1.848 = 3.3 cm3. Treating the source as a sphere, the radius corresponding to this volume is 0.925 cm, which is rounded to 1 cm. The density of this sphere is then 7g/(4r3/3 cm3) = 1.67 g/cm3.

The source is doubly encapsulated in stainless steel [13], although the encapsulation material is conservatively ignored. As the source may be in contact with the TMI-2 Fuel Canister inner shroud, the inner shroud is modeled as a 0.04-in thick steel shell in contact with the source (Page D.3 of [7], [26]).

This approximates what the neutron flux would be in close proximity to the source. The BORAL and outer shroud are conservatively ignored, and the Licon is modeled as a thin 0.197-in (0.5 cm) thick shell outside the inner shroud. The Licon, in general, is significantly thicker than 0.197-in at most locations, and the 0.197-in thickness is selected to maximize the fluence in the TMI-2 Fuel Canister shell, DSC shell, and HSM concrete. The TMI-2 Fuel Canister is modeled as a 0.25-in thick steel shell. The DSC is modeled as a shell 0.63-in thick in contact with the TMI-2 Fuel Canister. Finally, the concrete is modeled as 1 ft. thick at a distance of 3.9-in from the DSC to approximate the closest approach of the HSM to the DSC. The neutron fluence in the concrete is tallied over a 1 cm thickness on the inside of the HSM.

Because the water density within the Licon is not precisely known, four water densities are considered, consistent with the approach in the design basis source models.

Results are provided in Table 6.3-1. Given these conservative assumptions, the localized fluence is much higher than the fluence computed using the design basis source and the distributed source model.

However, the localized fluence remains below the limits. Results are maximized with twice the nominal water density, although the effect is somewhat negligible except within the Licon itself. The largest fluence in the steel occurs for the inner shroud, and the largest fluence in concrete occurs in the Licon.

This behavior is expected due to the proximity of these regions to the source.

The total neutron fluence (E > 0 MeV) in the shroud is 1.77x1015 n/cm2, which is below the fast fluence limit of 1018 n/cm2 (E > 1 MeV). The total neutron fluence (E > 0 MeV) in the Licon is 9.83x1014 n/cm2, which is below the fast fluence limit of 1019 n/cm2 (E > 1 MeV).

Note that these fluence values represent the maximum from a single TMI-2 Fuel Canister containing an AmBeCm source with spent fuel (8x106 n/s) rather than from 12 TMI-2 Fuel Canisters. However, the fluences for the AmBeCm source are so large that if the AmBeCm fluence values are added to the fluence values for 12 TMI-2 Fuel Canisters containing design basis sources as computed in Section 6.2, the total fluence is essentially the same because the AmBeCm fluence values are orders of magnitude CALC-3016518-002 Page 27 of 43

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Title:

Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal larger. Therefore, the AmBeCm results presented Table 6.3-1 are representative of an HSM loaded with 11 design basis TMI-2 Fuel Canisters and one TMI-2 Fuel Canister containing a mixture of fuel debris and an AmBeCm neutron source.

Table 6.3 AmBeCm Neutron Results Neutron Results (n/cm2)

Region Licon No Water Licon 1/2 Nominal Water Licon Nominal Water Licon 2x Nominal Water Maximum Shroud steel 1.73E+15 1.74E+15 1.75E+15 1.77E+15 1.77E+15 Licon 9.41E+14 9.51E+14 9.61E+14 9.83E+14 9.83E+14 TMI-2 Fuel Canister steel shell 4.93E+14 4.97E+14 5.00E+14 5.06E+14 5.06E+14 DSC steel 1.96E+14 1.96E+14 1.97E+14 1.98E+14 1.98E+14 HSM concrete 2.88E+13 2.89E+13 2.89E+13 2.90E+13 2.90E+13 CALC-3016518-002 Page 28 of 43

Page 29 of 43 Orano Federal Services

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal Figure 6.3 AmBeCm Source Model Source Inner Shroud Licon TMI-2 Fuel Canister DSC HSM CALC-3016518-002 Page 29 of 43

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal 7.0 Results/Conclusions The effect of neutron and gamma radiation on the stainless steel TMI-2 Canisters (Fuel, Knockout, and Filter), carbon steel DSC shell, and HSM concrete over a 60 year time period is investigated using MCNP. The effect of neutron and gamma radiation on the stainless steel TMI-2 Fuel Canister internals (stainless steel BORAL shroud and Licon) is also considered. The radiation effects on the BORAL shroud may be applied to other TMI-2 Canister steel internals, including TMI-2 Knockout and Filter Canister steel internals.

For the TMI-2 Canisters, the source decay from 1979 to 1985 is used to calculate the gamma dose accumulation during the 14-year period between approximately 1985 and 1999. Then, the gamma dose accumulation is calculated for a 60-year period starting in 1999 using a source decayed from 1979 to 1998. Summing these results produces the cumulative gamma dose accumulation. However, the neutron fluence accumulation is calculated over the 60 year period that includes time in which the TMI-2 Canisters are loaded with fuel but not loaded within the DSC or HSM (approximately 1985 to 1999).

This is because it was determined that the neutron source strength calculated from 1979 to 1998 is stronger than the neutron source strength calculated from 1979 to 1985. Gamma results are provided in Table 7.0-1 and neutron results are provided in Table 7.0-2. The results are well converged and the MCNP statistical uncertainty is small (<1%). The design basis source results for HSM concrete, TMI-2 Canister stainless steel shell, and DSC carbon steel shell are conservatively computed ignoring TMI-2 Canister internals. The results for TMI-2 Fuel Canister stainless steel internals applies to the steel internals of all TMI-2 Canister types. The Licon results are applicable only to the TMI-2 Fuel Canister, as the TMI-2 Knockout and Filter Canisters do not use this material. The results show acceptable margin to the limits. It is concluded that there are no adverse effects on these materials due to the additional 20 years of storage.

Two sets of neutron fluence results are provided: (1) using the design basis neutron source and (2) using the concentrated AmBeCm neutron source, which is approximately ~12x larger than the design basis neutron source. The neutron fluence may be significantly higher for the AmBeCm neutron source compared to the design basis source because it is larger in magnitude and smaller in volume, although the effect of this source would be highly localized and only resides within two TMI-2 Fuel Canisters.

Nevertheless, the estimated fluence values are several orders of magnitude below the limits, even for the AmBeCm source. The AmBeCm source neutron fluence results for HSM concrete, the TMI-2 Fuel Canister stainless steel shell, and DSC carbon steel shell include the Licon because the AmBeCm source is only in the TMI-2 Fuel Canister.

Note that for the neutron results, the total neutron fluence is computed (E > 0 MeV), while the limits are based upon fast fluence (E > 1.0 MeV). Therefore, there is additional conservatism in the results.

No formal literature searches are performed as part of this calculation.

CALC-3016518-002 Page 30 of 43

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal Table 7.0 Gamma Results, Design Basis Source Item Absorbed Energy (rad over 60 years)

Limit (rad)

Concrete Items TMI-2 Fuel Canister Licon 1.61E+09 1010 HSM Concrete 1.98E+08 Steel Items TMI-2 Canister Stainless Steel Internals 1.96E+09 3.8E+11 TMI-2 Canister Stainless Steel Shell 2.00E+09 None DSC Carbon Steel Shell 4.42E+08 Conservatively computed in model without internal hardware.

Computed in model with internal hardware.

The absorbed energy is rad over 60 + 14 = 74 years Table 7.0 Neutron Results Item Total Fluence (E > 0 MeV) (n/cm2 over 60 years),

Design Basis Source Total Fluence (E > 0 MeV) (n/cm2 over 60 years),

AmBeCm Source Fast Fluence Limit (E > 1 MeV)

(n/cm2)

Concrete Items TMI-2 Fuel Canister Licon 8.92E+11 9.83E+14 2E+19 HSM Concrete 5.49E+11 2.90E+13 Steel Items TMI-2 Canister Stainless Steel Internals 8.98E+11 1.77E+15 1018 TMI-2 Canister Stainless Steel Shell 9.23E+11 5.06E+14 DSC Carbon Steel Shell 5.67E+11 1.98E+14 Conservatively computed in model without internal hardware.

Computed in model with internal hardware.

CALC-3016518-002 Page 31 of 43

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal 8.0 Appendix A 8.1 References

1. MCNP5, MCNP - A General Monte Carlo N-Particle Transport Code, Version 5; Volume II:

Users Guide, LA-CP-03-0245, Los Alamos National Laboratory, April 2003.

2. Elleuch L.F., Dubois F., Rappeneau J., Paper SP 34-51, Effects of Neutron Radiation on Special Concretes and Their Components, Volume 34, pages 1071-1108, January 1, 1972, American Concrete Institute, Farmington Hills, MI.

3. NUREG-1611, Aging Management of Nuclear Power Plant Containments for License Renewal, US Nuclear Regulatory Commission, September 1997.

4. NUREG/CR-6927, Primer on Durability of Nuclear Power Plant Reinforced Concrete Structures - A Review of Pertinent Factors, Oak Ridge National Laboratory, February 2007.

5. CoC Renewal Application, Appendix 3E, Rev. 2, Evaluation of Neutron Fluence and Gamma Radiation on Storage System Structural Materials, June 2016.

6. General Public Utilities Nuclear, Inc. Three Mile Island Nuclear Station, GPU Nuclear Three Mile Island Nuclear Station Unit 2 Defueling Completion Report, April 11, 2011, NRC Accession Number ML111100641.

7. SAR-II-8.4, 2011, TMI-2 Safety Analysis Report, TOC-276, Rev. 29, Idaho Cleanup Project, February 2011.

8. Transnuclear West Drawing 219-02-2000, Rev. 1, Dry Shielded Canister Basket Assembly Safety Analysis Report.

9. Transnuclear West Drawing 219-02-2001, Rev. 1, Dry Shielded Canister Shell Assembly Safety Analysis Report.

10. Transnuclear West Drawing 219-02-6000, Rev. 1, Horizontal Storage Module Safety Analysis Report.

11. Babcock & Wilcox Drawing 1161300, Rev. 1, Fuel Canister SAR Information.

12. Department of Energy Letter from MD Gardner to JE Kaylor,

SUBJECT:

Approval of Licensing Evaluation TMl-01-002, Revision 1 (INTEC-NRC-01-019), March 27, 2000.

13. Engineering Design File EDF-1793, Rev. 4, Impact of AmBeCm Sources on the TMI-2 ISFSI Design Basis, Idaho Cleanup Project, March 28, 2001.

14. Vectra Calculation 219-02.0401, Rev. 0, INEL/TMI-2 Fuel Storage (Interim Storage System -

ISS).

15. AFS Document SD-MCNP-v5.1.51-CALC-3012586, Rev. 0, Software Dedication Report for MCNP v5.1.51 (Documentum Number SD-3012586-000).

16. AFS Document SD-SCALE-6.0-CALC-3012488, Rev. 0, Software Dedication Report for SCALE 6.0 (Documentum Number SD-3012488-000).

17. SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluations, ORNL/TM-2005/39, Version 6, Vols. I-III, January 2009.

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18. Nuclides and Isotopes, Fifteenth Edition, Chart of the Nuclides, General Electric Company and Knolls Atomic Power Laboratory, 1996.

19. Nuclear Wallet Cards, National Nuclear Data Center, Brookhaven National Laboratory, July 1990.

20. License SNM-2508, Amendment No.4, License for Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste, US Nuclear Regulatory Commission.

21. Vectra Calculation 219-02.0400, Rev. 0, Design Basis Source Term Calculation for the TMI-2 Debris Canisters.

22. EPRI Report 1015078, Plant Support Engineering: Aging Effects for Structures and Structural Components (Structural Tools), December 2007.

23. Babcock & Wilcox Drawing 1161301, Rev. 1, Knockout Canister SAR Information.

24. Babcock & Wilcox Drawing 1161299, Rev. 1, Filter Canister SAR Information.

25. Draft NUREG, Managing Aging Processess in Storage (MAPS) Report, US Nuclear Regulatory Commission, 2016.

26. Babcock & Wilcox Drawing 02-1095753E, Rev. 2, Fuel Canister Neutron Poison Shroud, Three Mile Island Unit 2.

27. Letter from JM Storton to PC Childress, Data Sheet Identifier 38-1013203-00, Ultralight Concrete, Babcock & Wilcox, January 24, 1985.

28. Idaho National Engineering and Environmental Laboratory Engineering Design File, EDF-1466, Rev. 2, Validation of Water Content in TMI-2 Canisters During Drying in the HVDS, June 28, 2000.

29. Idaho National Engineering and Environmental Laboratory Engineering Design File, EDF-797, Rev. 0, Water Ingress into TMI DSCs During Storage, March 2, 1999.

30. Standard Composition Library, ORNL/TM-2005/39, Version 6, Vol. III, Section M8, January 2009.

31. U.S. Nuclear Regulatory Commission, Federal Register Notices Publishing Environmental Assessments and Findings of No Significant Impacts For Requests For Exemptions From Requirements of 10 CFR Parts 20 and 72, March 13, 1999, Docket No. 72-20, NRC Accession Number 9903230211.

32. Department of Energy Letter from MD Gardner to JE Kaylor,

SUBJECT:

Transmittal of NRC Inspection Report 72-20/01-02 (INTEC-NRC-01-067), November 16, 2001.

33. John R. Lamarsh, Introduction to Nuclear Engineering, 2nd Edition, Addison Wesley, 1983.

34. LA-CP-13-00634, MCNP6TM Users Manual, Los Alamos National Laboratory, Version 1.0, May 2013, RSICC Package ID C00834MNYCP02

35. AREVA Federal Services Calculation, SD-3012588, Software Dedication Report for MCNP v6.1.00, Revision 0.

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal

36. NRC Correspondence, Request for Additional Information for the Technical Review of the Application for Renewal of the Three Mile Island Unit 2 Independent Spent Fuel Storage Installation License No. SNM-2508, January 29, 2018, NRC Accession Number ML18030A172.

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal 8.2 Computer Software Usage 8.2.1 Computer Software Usage for Rev. 0 8.2.1.1 Installation and In-Use Testing for SCALE SCALE6 is executed on the same computer used for the validation activities documented in [16]. Therefore, installation testing is not required. For in-use testing, test case Co60_RM1.inp from [16] is renamed Co60_RM1c.inp and run on computer RMIGLIORE1. Other than date/time stamp differences, the results are identical to the equivalent results from [16], indicating the in-use testing is successful.

8.2.1.2 File Listing for SCALE COMPUTER RUN RECORD Run description See input/output file names and the discussions in the body of the calculation Software used SCALE6.0 Computer name RMIGLIORE1, QA verification for SCALE 6.0 has been performed for this computer [16].

Hardware (processor name)

Intel Xeon CPU E5-1650 0 @ 3.20 GHz Operating system 64-bit Windows 7 Enterprise Unique run identifier See input/output file names and associated date/time stamps List of input/output files See input/output file names Basis supporting the application of the computer program to the specific physical problem being analyzed The basis for using SCALE6.0 for source term development is documented in [16].

For input files, date/time stamps are generated using the DOS DIR command on a Windows PC in the Pacific time zone. Windows Explorer sometimes rounds the minute differently than DOS DIR, so time/date stamps should only be compared using the DOS DIR command. Also, the time can be listed either as Pacific Daylight Savings or Pacific Standard time, depending on when the list was generated. Because Windows and/or DOS DIR adjusts the reported time depending on the time of year, 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> differences in the time stamps have been observed due to changes between Standard and Daylight Savings time (e.g., if the list is generated in Standard time but the list is verified in Daylight Savings time, the times will be different by 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />). Also, the observed time stamp could be different if the files are ultimately stored on a computer in a different time zone CALC-3016518-002 Page 35 of 43

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Title:

Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal (i.e., if the files are stored on a computer in the Eastern time zone the time stamps will shift 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> because the listed times represent Pacific time).

This behavior is acceptable.

For output files, which contain a time stamp internal to the output file, the internal time stamp is provided rather than the DOS DIR time stamp. As each output file has multiple time stamps, the time stamp in the job started line is used.

Input Date Time Output Date Time Co60_RM1c.inp 7/24/2012 1:45 PM Co60_RM1c.out 8/17/2016 9:42:27.32 source.inp 8/22/2016 1:36 PM source.out 8/22/2016 13:36:34.04 8.2.2 Computer Software Usage for Rev. 1 8.2.2.1 Installation and In-Use Testing for MCNP Installation testing is recommended for MCNP5 v1.51 on computer RMIGLIORE1 because the validation activities were performed on computer MMACQUIGG1 [15]. Per Section 8 of the MCNP software dedication report [15], the following three items should be verified: (1) the correct cross-section data files are installed, (2) the correct executables are installed, and (3) the hardware/software environment matches the validation computer. Items (1) and (2) are verified by comparing the file sizes of the data libraries and executables between computers MMACQUIGG1 and RMIGLIORE1. This activity in documented in Excel spreadsheet INSTALLATION.XLSX. All file sizes match. Item (3) is verified by inspection, as machines RMIGLIORE1 and MMACQUIGG1 both use the Windows 7 Enterprise operating system with Intel Xeon processors. The type of Xeon processor is not identical between the two machines, although this is acceptable. Therefore, installation testing for machine RMIGLIORE1 is acceptable.

In-use testing is also required. Test case buildup_pb_5_mpf.i from the MCNP software dedication report [15] is renamed buildup_pb_5_mpf_RM1h.i and run on computer RMIGLIORE1. Other than differences due to run-time, the results are identical to the equivalent results from the MCNP software dedication report [15], indicating the in-use testing is successful. A comparison of the tally results are provided in the following table.

Original

In use 2.52589E-08 0.0026 2.52589E-08 0.0026 3.61964E-08 0.0018 3.61964E-08 0.0018 6.14553E-08 0.0018 6.14553E-08 0.0018 CALC-3016518-002 Page 36 of 43

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal 8.2.2.2 File Listing for MCNP COMPUTER RUN RECORD Run description See input/output file names and the discussions in the body of the calculation Software used MCNP5 v1.51 Computer name RMIGLIORE1, QA verification for MCNP5 has been performed for computer MMACQUIGG1 [15]. Installation testing is required.

Hardware (processor name)

Intel Xeon CPU E5-1650 0 @ 3.20 GHz Operating system 64-bit Windows 7 Enterprise Unique run identifier See input/output file names and associated date/time stamps List of input/output files See input/output file names Basis supporting the application of the computer program to the specific physical problem being analyzed The basis for using MCNP5 shielding analysis documented in [15].

For input files, date/time stamps are generated using the DOS DIR command on a Windows PC in the Pacific time zone. Windows Explorer sometimes rounds the minute differently than DOS DIR, so time/date stamps should only be compared using the DOS DIR command. Also, the time can be listed either as Pacific Daylight Savings or Pacific Standard time, depending on when the list was generated. Because Windows and/or DOS DIR adjusts the reported time depending on the time of year, 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> differences in the time stamps have been observed due to changes between Standard and Daylight Savings time (e.g., if the list is generated in Standard time but the list is verified in Daylight Savings time, the times will be different by 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />). Also, the observed time stamp could be different if the files are ultimately stored on a computer in a different time zone (i.e., if the files are stored on a computer in the Eastern time zone the time stamps will shift 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> because the listed times represent Pacific time).

This behavior is acceptable.

For output files and associated tally files, which contain a time stamp internal to the output file, the internal time stamp is provided rather than the DOS DIR time stamp.

Input Date Time Output Tally Date Time TMIg_mk3.i 10/28/2016 2:09 PM TMIg_mk3.o TMIg_mk3.m 10/28/2016 14:10:13 TMIn_mk3.i 10/28/2016 2:13 PM TMIn_mk3.o TMIn_mk3.m 10/28/2016 14:14:04 AmBeCm_licon_nom.i 10/31/2016 7:20 AM AmBeCm_licon_nom.o AmBeCm_licon_nom.m 10/31/2016 7:20:57 CALC-3016518-002 Page 37 of 43

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal AmBeCm_licon_w0.i 10/31/2016 7:20 AM AmBeCm_licon_w0.o AmBeCm_licon_w0.m 10/31/2016 7:22:14 AmBeCm_licon_wdouble.i 10/31/2016 7:20 AM AmBeCm_licon_wdouble.o AmBeCm_licon_wdouble.m 10/31/2016 7:23:32 AmBeCm_licon_whalf.i 10/31/2016 7:20 AM AmBeCm_licon_whalf.o AmBeCm_licon_whalf.m 10/31/2016 7:24:49 TMIg_licon_nom.i 10/31/2016 6:58 AM TMIg_licon_nom.o TMIg_licon_nom.m 10/31/2016 6:59:13 TMIg_licon_w0.i 10/31/2016 6:58 AM TMIg_licon_w0.o TMIg_licon_w0.m 10/31/2016 7:00:23 TMIg_licon_wdouble.i 10/31/2016 6:58 AM TMIg_licon_wdouble.o TMIg_licon_wdouble.m 10/31/2016 7:01:33 TMIg_licon_whalf.i 10/31/2016 6:58 AM TMIg_licon_whalf.o TMIg_licon_whalf.m 10/31/2016 7:02:40 TMIn_licon_nom.i 10/31/2016 6:58 AM TMIn_licon_nom.o TMIn_licon_nom.m 10/31/2016 7:03:55 TMIn_licon_w0.i 10/31/2016 6:58 AM TMIn_licon_w0.o TMIn_licon_w0.m 10/31/2016 7:05:23 TMIn_licon_wdouble.i 10/31/2016 6:58 AM TMIn_licon_wdouble.o TMIn_licon_wdouble.m 10/31/2016 7:06:38 TMIn_licon_whalf.i 10/31/2016 6:58 AM TMIn_licon_whalf.o TMIn_licon_whalf.m 10/31/2016 7:07:54 buildup_pb_5_mfp_RM1h.i 4/16/2015 6:25 PM buildup_pb_5_mfp_RM1h.o N/A 10/26/2016 10:53:29 8.2.3 Computer Software Usage for Rev. 2 Computer Name:

SGIBBONEY5 Hardware Profile of Computer:

Intel Xeon CPU E5-1650 @ 3.60 GHz, 32.0 GB RAM Operating System:

64-bit Windows 7 Enterprise, Service Pack 1 8.2.3.1 In-Use Testing for MCNP Input file buildup_pb_5_mpf.i is taken from the MCNP6.1.00 software dedication report [35] for in-use testing. The file is run on 5/11/2018. The results in the output file is identical to that in [35] except for run-unique identifiers (such as date and time of run), indicating that MCNP6.1 performs as expected and is acceptable for use.

CALC-3016518-002 Page 38 of 43

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal 8.2.3.2 File Listing for MCNP Directory: IO_Rev2\\Appendix B Mode LastWriteTime Length Name

-a--- 8/2/2018 3:40 PM 5613 TMIg_licon_w0_40.i

-a--- 8/2/2018 3:41 PM 6435 TMIg_licon_w0_40.m

-a--- 8/2/2018 3:41 PM 70382 TMIg_licon_w0_40.o

-a--- 8/2/2018 3:31 PM 4373 TMIg_mk3_40.i

-a--- 8/2/2018 3:32 PM 5898 TMIg_mk3_40.m

-a--- 8/2/2018 3:32 PM 65478 TMIg_mk3_40.o Directory: IO_Rev2\\DBS Mode LastWriteTime Length Name

-a--- 5/10/2018 5:33 PM 5644 TMIg_licon_nom_6.i

-a--- 5/11/2018 3:28 PM 6435 TMIg_licon_nom_6.m

-a--- 5/11/2018 3:28 PM 70617 TMIg_licon_nom_6.o

-a--- 5/10/2018 5:34 PM 5619 TMIg_licon_w0_6.i

-a--- 5/11/2018 3:29 PM 6435 TMIg_licon_w0_6.m

-a--- 5/11/2018 3:29 PM 70488 TMIg_licon_w0_6.o

-a--- 5/10/2018 5:36 PM 5644 TMIg_licon_wdouble_6.i

-a--- 5/11/2018 3:31 PM 6435 TMIg_licon_wdouble_6.m

-a--- 5/11/2018 3:31 PM 70629 TMIg_licon_wdouble_6.o

-a--- 5/10/2018 5:37 PM 5644 TMIg_licon_whalf_6.i

-a--- 5/11/2018 3:32 PM 6435 TMIg_licon_whalf_6.m

-a--- 5/11/2018 3:32 PM 70597 TMIg_licon_whalf_6.o

-a--- 5/10/2018 5:38 PM 4376 TMIg_mk3_6.i

-a--- 5/11/2018 3:34 PM 6106 TMIg_mk3_6.m

-a--- 5/11/2018 3:34 PM 61241 TMIg_mk3_6.o Directory: IO_Rev2\\In-Use Testing CALC-3016518-002 Page 39 of 43

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Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal Mode LastWriteTime Length Name

-a--- 5/11/2018 3:10 PM 2974 buildup_pb_5_mfp.i

-a--- 5/11/2018 3:26 PM 26456 buildup_pb_5_mfp.o Directory: IO_Rev2\\Spreadsheet Mode LastWriteTime Length Name

-a--- 5/14/2018 11:06 AM 60772 Results_R2.xlsx 8.2.4 Excel 2010 File Listing Excel is used to process data. The following spreadsheets are used in the calculation.

Spreadsheet Purpose DATA_R1.XLSX Collection of various input data RESULTS_R1.XLSX Summary of results RESULTS_R2.XLSX Summary of results INSTALLATION.XLSX Data size comparison to verify the MCNP5 installation CALC-3016518-002 Page 40 of 43

Page 41 of 43 Orano Federal Services

Title:

Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal 9.0 Appendix B To provide the technical basis for Orano Federal Services response to RAI 3-11 of Reference [36], this appendix determines the dose rates experienced by the TMI-2 Canisters and the stainless steel subcomponents at the beginning of the Period of Extended Operations (PEO in 2019). There is a potential that a radiation-induced localized corrosion mechanism may exist within the TMI-2 Canister internal environment. The mechanism for radiation-induced localized corrosion is due to the presence of gamma radiation and its interaction with water within the TMI-2 Canisters. Such an aging mechanism could possibly generate radiolytic oxidizing byproducts (e.g., hydrogen peroxide, nitric acid). The presence of these byproducts could increase the corrosion potential and therefore could propagate existing corrosion processes (e.g. pitting, crevice, or stress corrosion cracking). Therefore, the calculated gamma dose rates experienced by the TMI-2 Canisters, including internals, has been requested in support of the determination of whether such a corrosion mechanism could compromise the intended function of stainless steel subcomponents inside the TMI-2 Canisters.

The main analysis was concerned with calculating the cumulative dose; however this appendix is interested in calculating the dose rates at the beginning of the PEO. To do that, the 40-year source term is obtained from source.out to determine the intensity of the source term at that time. The design basis gamma source term at 40 years post-accident (decayed from 1979 to 2019) is shown in Table 9.0-1. This 40-year source term is used in the same MCNP models as in the main analysis; however modifications are made to the models to calculate the energy deposition in rad per hour by the gamma source into the TMI-2 Canisters and the internal components.

The absorbed energy is computed by MCNP in units of MeV/(g-s) and converted to rad/s by multiplying by the conversion factor 1 MeV/g = 1.602 x 10-8 rad. The total absorbed energy is then obtained by multiplying by the total time (in seconds). To calculate the per-hour rad amount experienced by the components, a conversion factor of (1.602x10-8 rad/(MeV/g)) x (3600 s/hr) = 5.7672x10-5 rad-s-g/MeV-hr is applied in MCNP F6 tallies (as the default unit is MeV/(g-s)) via the FM cards.

The design basis gamma source in 2019 is provided on a per-TMI-2 Canister basis. Because the MCNP models from the main analysis require the total source inside the DSC, and there are up to 12 TMI-2 Canisters inside each DSC, the total gamma source input is 2.754x1014 x 12 = 3.305x1015 /s per DSC.

The material properties for the fuel, stainless steel, and concrete used in the MCNP models in the main analysis remain applicable for the analysis performed in this appendix.

CALC-3016518-002 Page 41 of 43

Page 42 of 43 Orano Federal Services

Title:

Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal Table 9.0 Design Basis Gamma Source Term (2019)

Eupper (MeV)

Source per TMI-2 Canister

(/s) 5.00E-02 8.351E+13 1.00E-01 2.587E+13 2.00E-01 1.534E+13 3.00E-01 4.849E+12 4.00E-01 3.447E+12 6.00E-01 2.328E+12 8.00E-01 1.391E+14 1.00E+00 5.184E+11 1.33E+00 4.004E+11 1.66E+00 5.944E+10 2.00E+00 8.795E+09 2.50E+00 4.510E+08 3.00E+00 4.275E+06 4.00E+00 4.810E+04 5.00E+00 1.596E+04 6.50E+00 6.312E+03 8.00E+00 1.222E+03 1.00E+01 2.571E+02 Totals 2.754E+14 TMI-2 Canister without Internal Hardware In this model, the fuel with design basis sources is conservatively in contact with the wall of the TMI-2 Canister, as was done in the first set of MCNP models in the main analysis. In the main analysis, the TMI-2 Canister without Internal Hardware model produced bounding results for the energy deposition in the canister shell in comparison to the model with internal hardware modeled. The gamma energy deposition tallies remain in the same setup as in the main analysis.

The computed energy deposition for the average canister is 1520 rad/hr with the maximum individual canister being 1745 rad/hr.

TMI-2 Fuel Canister with Internal Hardware In the second set of MCNP models in this appendix, the internals of the TMI-2 Fuel Canister are modeled to determine the dose rate in rad/hr experienced by the stainless steel shroud. Again, the gamma energy deposition tallies remain in the same setup as in the main analysis.

The computed energy deposition in the shroud for the average canister with no water content in the Licon, as no water in the Licon was shown to yield the highest dose rates of all Licon water contents CALC-3016518-002 Page 42 of 43

Page 43 of 43 Orano Federal Services

Title:

Radiation Effects on Materials for TMI-2 ISFSI License Renewal Doc./Rev.: 02029.00.0000.02-01, Rev. 2 Project: 02029.00.0000.02, NRC License Renewal analyzed in Table 6.2-3, is 1693 rad/hr. The maximum individual shroud with nominal water is 1680 rad/hr.

CALC-3016518-002 Page 43 of 43