ML100560175

Independent Spent Fuel Storage Installation - Response to Request for Additional Information for License Amendment Request No. 9

ML100560175
Person / Time
Site:	Calvert Cliffs
Issue date:	02/18/2010
From:	George Gellrich Constellation Energy Nuclear Group, EDF Group
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards
References
TAC L24350
Download: ML100560175 (58)

Text

George H. Gellrich Calvert Cliffs Nuclear Power Plant, LLC Vice President 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 410.495.5200 410.495.3500 Fax CENG a joint venture of l Cornstellatilon <'eDF w Energyý:;,D CALVERT CLIFFS NUCLEAR POWER PLANT February 18, 2010 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Document Control Desk

SUBJECT:

Calvert Cliffs Nuclear Power Plant Independent Spent Fuel Storage Installation; Docket No. 72-8 Response to Request for Additional Information for License Amendment Request No. 9 (TAC No. L24350)

REFERENCES:

(a) Letter from Mr. J. A. Spina (CCNPP) to Document Control Desk (NRC),

dated June 15, 2009, License Amendment Request: Allow Increased Burnup Fuel to be Loaded into NUHOMS-32P Dry Shielded Canister (b) Letter from Mr. J. Goshen (NRC) to Mr. J. A. Spina (CCNPP), dated January 21, 2010, First Request for Additional Information for License Amendment Request No. 9 to Materials License No. SNM-2505, Calvert Cliffs Independent Spent Fuel Storage Installation (TAC No. L24350)

Calvert Cliffs Nuclear Power Plant, LLC submitted a license amendment request (Reference a) to allow the loading of increased burnup fuel into a NUHOMS-32P Dry Shielded Canister. The Nuclear Regulatory Commission (NRC) staff has determined that additional information is needed to complete their review (Reference b). Attachment (1) and the enclosed CD provide the requested information.

Enclosures (1) and (2) contain the requested calculations. The response to one request for additional information resulted in a change to the proposed Technical Specifications pages previously submitted in Reference (a). The marked up pages for the proposed change are contained in Attachment (2). The pages in Attachment (2) are in addition to and supersede the same pages previously submitted in Reference (a).

These additional changes to the Technical Specifications do not significantly change the environmental assessment provided in Reference (a) and the categorical exclusion set forth in 10 CFR 51.22(c)( 11) is still valid.

Document Control Desk February 18, 2010 Page 2 Should you have questions regarding this matter, please contact Mr. Douglas E. Lauver at (410) 495-5219.

Very truly yours, STATE OF MARYLAND

TO WIT:

COUNTY OF CALVERT, I, George H. Gellrich, being duly sworn, state that I am Vice President - Calvert Cliffs Nuclear Power Plant, LLC (CCNPP), and that I am duly authorized to execute and file this License Amendment Request on behalf of CCNPP. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other CCNPP employees and/or consultants. Such information has been reviewed in accordance with company practice and I believe it to be reliable.

Subscribed and sworn before me, a Notary Public in and for the State of Maryland and County of IYPIA' w1 ,o this /&'/J'-day of , -, c/ , 2010.

V.TI.-ESVTSNE$,Tn' H ild~and Notarial Seal: Notary Public e1426V My Commission Expires:

(I bate GHG/PSF/bjd Attachments: (1) Response to Request for Additional Information (and CD included)

Enclosures:

(1) Transnuclear Calculation 1095-577 (2) Transnuclear Specification E- 18851, Rev. 7 (2) Marked Up Technical Specification Pages

Document Control Desk February 18, 2010 Page 3 cc: D. V. Pickett, NRC E. W. Brach, NRC S. J. Collins, NRC J. Goshen, NRC Resident Inspector, NRC M. F. Weber, NRC S. Gray, DNR

ATTACHMENT (1)

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION Calvert Cliffs Nuclear Power Plant, LLC February 18, 2010

ATTACHMENT (1)

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION Calvert Cliffs Nuclear Power Plant, LLC (CCNPP) submitted a license amendment request (Reference 1) to allow the loading of increased burnup fuel into a NUHOMS-32P Dry Shielded Canister. The Nuclear Regulatory Commission (NRC) staff has completed its initial review and determined that additional information is needed (Reference 2). The responses to the NRC staffs request for additional information are presented below.

CHAPTER 4.0 THERMAL EVALUATION RAI 4-1:

Provide the supporting calculation packages and specifications listed as references to "Thermal Analysis of NUHOMS-32P+ DSC for Vacuum Drying Condition," Document No. NUH32P+.0401, along with the ANSYS input files that support this calculation and the results given within the SAR. 1, "Transnuclear, Inc. Calculation, "Thermal Analysis of NUHOMS 32P+ DSC for Vacuum Drying Condition," Document No. NUH32P+.0401" refers to two calculation packages: "Thermal Analysis of Vacuum Drying, Calculation No. 1095-57, Rev. 0" and "Design Criteria for the NUHOMS-32P Storage System for Calvert Cliff Nuclear Plant, Specification No. E-18851, Rev. 7."

Constellation Energy should provide these two documents, along with the ANSYS input data files used within the analysis in order for staff to confirm that cladding limits are being met for vacuum drying operations.

This information is needed to confirm compliance with 10 CFR 72.122(h)(1) and 10 CFR 72.122(1).

RAI 4-1 Response:

The supporting electronic files for Transnuclear calculation NUH32P+.0401 are included on the attached CD-ROM. A copy of Transnuclear calculation 1095-57 is included as Enclosure (1) and the associated electronic files are also included on the attached CD-ROM. A copy of Transnuclear specification E-18851 Rev. 7 is included as Enclosure (2).

RAI 4-2:

Provide a discussion of the impact, including revised calculation results or a sensitivity study, of reduced thermal conductivity of high bum up fuel cladding on the effective thermal conductivity of the fuel calculated as part of the SAR analysis.

It is understood by the staff that there can be a decrease of up to 50% in the thermal conductivity of the fuel cladding for assemblies with burnups of greater that 45 GWD/MTU. This effect needs to be addressed in order for the staff to make an assessment of the ability of the fuel cladding to meet the performance requirements in 10 CFR Part 72.

This information is needed to confirm compliance with 10 CFR 72.122(h)(1) and 10 CFR 72.122(l).

RAI 4-2 Response:

We reviewed the Westinghouse Topical Report CENPD-404-P-A, "Implementation of ZIRLO Cladding Material in CE Nuclear Power Fuel Assembly Designs." This report was approved by the NRC staff for use at Calvert Cliffs Units 1 and 2 with a burnup limit of 60 GWd/MTU in License Amendment 251/228 (see ML020790273). This topical report provides a summary of the models of Zircaloy-4 metal thermal conductivity used in fuel performance and ECCS evaluation codes licensed for use at Calvert Cliffs.

These models show no dependence of Zircaloy thermal conductivity on burnup.

I

ATTACHMENT (1)

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION However, it is also recognized that the buildup of lower thermal conductivity oxide on the rod water side with increased burnup (up to a maximum of 125 microns for a burnup of 52 GWd/MTU) can potentially result in higher cladding temperatures at the metal/oxide interface. To address the potentially adverse effect of cladding oxidation on fuel rod temperature, two temperature gradients are considered:

A) temperature gradient across the fuel pellet, and B) temperature gradient across the rod.

For item A, the temperature gradient across the fuel pellet is driven solely by linear heat generation rate and by thermal conductivity of uranium dioxide. For item B, the temperature gradient across the rod is further affected by the heat transfer coefficient at the cladding outer surface, heat transfer coefficient at the gap region, and cladding thermal conductivity. Any oxidation of the cladding does affect pellet surface, pellet centerline, as well as peak cladding temperature. However, the increase in these temperatures is insignificant.

To demonstrate that cladding oxidation has an insignificant effect on peak fuel pellet and peak cladding temperatures, a typical fuel rod at an average linear heat generation rate of 7 kW/ft is analyzed. The thickness of the oxide layer is a function of primary side chemistry control and burnup among other factors. In this analysis, a parametric study was performed in which the thickness of oxide layer was increased by increments of 10 micron, from no oxide layer to a thickness of 200 microns. Thermal conductivity of the cladding from various sources was estimated to be on the order of 16 W/m K (9 Btu/hr ft 'F). The oxide conductivity is degraded down to about 10% of the cladding conductivity, which is on the order of 2 W/m K (1 Btu/hr ft 'F).

The analysis indicates that there is a 0.2% increase in the peak pellet temperature for every 10 micron increase in the cladding oxidation under reactor operating conditions. As shown in the below figure, this ratio remains nearly constant for thickness of the oxidation layer ranging from 0 to 200 microns.

2940 2920 2900 CL E 2880 I-2860 i 2840 0- 2820 2800 0 50 100 150 200 Oxide Layer Thickness (micron)

At the bounding oxide thickness for 52 GWd/MTU these results are consistent with those cited on page 7 of CEN-382(B)-P, "Verification of the Acceptability of a 1-Pin Burnup Limit of 60 MWd/kg for Calvert Cliffs Units 1 and 2," which the NRC staff approved on July 16, 1992 (TAC No. M74169/M74170). This 2

ATTACHMENT (1)

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC Safety Evaluation Report based its approval on a previous Safety Evaluation Report for CE 16x1 6 fuel issued on June 22, 1992 (TAC No. M82192) which stated:

"The upper bound oxide thickness at a rod-average burnup of 60MWd/kgM was used to estimate the increase in cladding temperatures and stress, and found to have little impact on either of these analyses. Therefore, we conclude that cladding oxidation is acceptable for the CE 16xl6 fuel design in ANO-2 up to a rod-average bumup of 60 MWd/kgM."

For dry storage application, the fuel assemblies loaded in the NUHOMS-32P canisters have a maximum design limit heat generation rate of 0.66 kW/assembly. This amounts to 0.33 W/ft (considering 176 rods per assembly and a heated length of 11.4 ft). When this case was analyzed, the rise in peak pellet temperature as well as the peak cladding temperature (°F) was 0.03% for an oxide thickness of 200 micron. This oxide thickness bounds the upper limit for an assembly with an average burnup 52 GWd/MTU and would produce about a 1/2/ 2'F increase at the ISG- II cladding temperature limit of 752°F. There is more than enough margin indicated in Table 7-2 of Transnuclear calculation NUH32P+.0401 to accommodate such an increase.

CHAPTER 5.0 SHIELDING EVALUATION RAI 5-1:

Justify the changes to Technical Specification (TS ) 2.1 that establish new neutron and gamma source term limits allowed in each fuel assembly.

Interim Staff Guidance (ISG)-6 states, "Absent adequate justification acceptable to the staff, the SAR should not attempt to establish specific source terms as operating controls and limits for cask use." The staff believes that it may be more appropriate to eliminate this technical specification altogether and then rely on limiting the maximum assembly bumup, cooling time, enrichment, and decay heat as this methodology is the standard currently used by other applicants. The staff requests Constellation Energy evaluate this option and provide its response.

This information is needed to ensure that the storage system continues to meet the extemal dose rate requirements of 10 CFR 72.104 and 72.106.

RAI 5-1 Response:

The changes proposed to Technical Specification 2.1 for this License Amendment Request were modeled after those previously approved by the NRC for Independent Spent Fuel Storage Installation (ISFSI)

License Amendment 6 on October 6 h, 2005 (see ML051010396). Technical Specification 2.1 is currently met in our fuel loading procedures for the NUHOMS-32P by requiring that fuel selected for loading also meet the minimum required cooling time in CCNPP calculation CA06432, "32P Assembly Insertion Requirements." This calculation was also previously submitted to the NRC (see ML041380206). It was intended that CCNPP calculation CA06721 (Attachment 4 of Reference 1) would fulfill the same role as CA06432 currently does to ensure that Technical Specification 2.1 is met for fuel selected for loading.

We would prefer to maintain the current format for Technical Specification 2.1 proposed in Reference I to avoid altering the licensing basis of the previous 63 NUHOMS-24P and NUHOMS-32P DSCs loaded at CCNPP.

Technical Specification 3.1.1(6) requires that fuel selected for loading meet the cooling times specified in ISFSI Updated Safety Analysis Report (USAR) Table 9.4.1. To provide additional assurance that Technical Specification 2.1 is met, we propose adding the NUHOMS-32P specific fuel qualification table 3

ATTACHMENT (1)

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION shown below to ISFSI USAR Table 9.4.1. This table was created from the fit of time to cool to 660 watts as a function of assembly enrichment and burnup provided in Section 6.2 of CA0672 1. As described in CA06721, an assembly neutron source of 4.175E8 neutrons/sec per assembly was selected for use in the shielding analyses based on a review of the Calvert Cliffs spent fuel pool inventory of standard CE 14x14 fuel. Section 6.2 of CA06721 also provides a fit of assembly neutron source strength at the time it has cooled to 660 watts as a function of enrichment and burnup. This fit has been used to shade regions where additional time beyond that necessary to cool to 660 watts is required to ensure that the neutron source remains below that used in the shielding analyses. A lower bound insertion time of 7 years has also been set in the table. A note at the bottom of the table would indicate that fuel in these shaded regions would require an assembly specific source term calculation to determine the additional cooling time required to meet Technical Specification 2.1 to be eligible for loading. As can be seen in Table 2-7 of CA06721, there is only one standard CE 14x14 assembly in the CCNPP spent fuel pool that would fall into the shaded region.

Post-Discharge Cooling Time (years) to Meet 660W Decay Heat and 4.175E8 n/sec per Assembly for NUHOMS-32P (Proposed Addition to USAR Table 9.4.1 for NUHOMS-32P)

Bumup (GWd/ 38< 39< 40< 41< 42< 43< 44< 45< 46< 47< 48< 49< 50< 51<

MTU) B< B B B B B B B B B B B B B B 38 <39 *40 *41 *42 *43 *44 *45 <46 *47 *<48 _<49 _<50 *51 *52 Enrichment 1 2.00<E<2.10 8.4 8.9 9.4+ 9.9+ 10.5+ 11.2+ 11.9+ 12.6+ 13.4+ 14.3+ 15.2+ 16.2+ 17.3+ 18.5+ 19.7+

2.10_<E<2.20 8.3 8.7 9.2 9.8+ 10.4+ 11.0+ 11.7+ 12.5+ 13.3+ 14.1+ 15.1+ 16.0+ 17.1+ 18.2+ 19.4+

2.20<E<2.30 8.1 8.6 9.1 9.6 10.2+ 10.8+ 11.5+ 12.3+ 13.1+ 13.9+ 14.9+ 15.9+ 16.9+ 18.0+ 19.2+

2.30<E<2.40 8.0 8.5 9.0 9.5 10.1+ 10.7+ 11.4+ 12.1+ 12.9+ 13.8+ 14.7+ 15.7+ 16.7+ 17.8+ 19.0+

2.40<E<2.50 7.9 8.4 8.8 9.4 9.9 10.6+ 11.2+ 12.0+ 12.8+ 13.6+ 14.5+ 15.5+ 16.5+ 17.7+ 18.8+

2.50<E<2.60 7.8 8.2 8.7 9.3 9.8 10.4 11.1+ 11.8+ 12.6+ 13.5+ 14.4+ 15.3+ 16.4+ 17.5+ 18.7+

2.60<E<2.70 7.7 8.2 8.6 9.1 9.7 10.3 11.0 11.7+ 12.5+ 13.3+ 14.2+ 15.2+ 16.2+ 17.3+ 18.5+

2.70<E<2.80 7.6 8.1 8.5 9.0 9.6 10.2 10.9 11.6 12.4+ 13.2+ 14.1+ 15.0+ 16.1+ 17.2+ 18.3+

2.80<E<2.90 7.6 8.0 8.4 9.0 9.5 10.1 10.8 11.5 12.2 13.1+ 13.9+ 14.9+ 15.9+ 17.0+ 18.2+

2.90<E<3.00 7.5 7.9 8.4 8.9 9.4 10.0 10.7 11.4 12.1 12.9 13.8+ 14.8+ 15.8+ 16.9+ 18.0+

3.00<E<3.10 7.4 7.9 8.3 8.8 9.3 9.9 10.6 11.3 12.0 12.8 13.7 14.6+ 15.7+ 16.7+ 17.9+

3.10<E<3.20 7.4 7.8 8.2 8.7 9.3 9.8 10.5 11.2 11.9 12.7 13.6 14.5 15.5+ 16.6+ 17.7+

3.20<E<3.30 7.3 7.7 8.2 8.7 9.2 9.8 10.4 11.1 11.8 12.6 13.5 14.4 15.4 16.5+ 17.6+

3.30<E<3.40 7.3 7.7 8.1 8.6 9.1 9.7 10.3 11.0 11.7 12.5 13.4 14.3 15.3 16.4 17.5 3.40<E<3.50 7.3 7.6 8.1 8.5 9.1 9.6 10.2 10.9 11.6 12.4 13.3 14.2 15.2 16.2 17.4 3.50<E<3.60 7.2 7.6 8.0 8.5 9.0 9.6 10.2 10.8 11.6 12.3 13.2 14.1 15.1 16.1 17.2 3.60<E<3.70 7.2 7.6 8.0 8.4 8.9 9.5 10.1 10.8 11.5 12.3 13.1 14.0 15.0 16.0 17.1 3.70<E<3.80 7.2 7.5 7.9 8.4 8.9 9.4 10.0 10.7 11.4 12.2 13.0 13.9 14.9 15.9 17.0 3.80<E<3.90 7.1 7.5 7.9 8.3 8.8 9.4 10.0 10.6 11.3 12.1 12.9 13.8 14.8 15.8 16.9 3.90-<E<4.00 7.1 7.5 7.9 8.3 8.8 9.3 9.9 10.5 11.2 12.0 12.8 13.7 14.7 15.7 16.8 4.00<E<4.10 7.1 7.4 7.8 8.2 8.7 9.3 9.8 10.5 11.2 11.9 12.7 13.6 14.6 15.6 16.7 4.10<E<4.20 7.0 7.4 7.8 8.2 8.7 9.2 9.8 10.4 11.1 11.8 12.7 13.5 14.5 15.5 16.6 4.20<E<4.30 7.0 7.4 7.7 8.2 8.6 9.1 9.7 10.3 11.0 11.8 12.6 13.4 14.4 15.4 16.5 4.30:5E<4.40 7.0 7.3 7.7 8.1 8.6 9.1 9.7 10.3 10.9 11.7 12.5 13.3 14.3 15.3 16.4 4.40<E<4.50 7.0 7.3 7.6 8.1 8.5 9.0 9.6 10.2 10.9 11.6 12.4 13.2 14.2 15.2 16.2

+ indicates that additional cooing time beyond tnat shown must be determined tnrougn an assembly specitic source term calculation to ensure compliance with Technical Specification 2.1.

4

ATTACHMENT (1)

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION CHAPTER 8.0 MATERIALS EVALUATION RAI 8-1:

The staff requests Constellation Energy evaluate the removal of references to utilization of air during the blow-down of spent fuel from the amendment request along with specifying in the TS that the blow-down of the spent fuel will be done with an inert gas.

The TS permit loading of fuel with pinhole leaks and larger defects (permitted that such defects do not adversely affect fuel handling and transfer). The exposure of spent fuel with pinhole leaks, hairline cracks, or other breaches in the cladding is prohibited due to the potential for oxidation of the fuel pellets and subsequent rod splitting.

This information is needed to evaluate compliance with 72.122(h) & (1).

RAI 8-1 Response:

Interim Staff Guidance-22, Revision 2, "Potential Rod Splitting Due to Exposure to an Oxidizing Atmosphere During Short-Term Cask Loading Operations in LWR or Other Uranium Oxide Based Fuel,"

provides the following three options to address the potential for and consequences of fuel oxidation:

1. Maintain the fuel rods in an appropriate environment such as Ar, N2, or He to prevent oxidation;

2. Assure that there are not any cladding breaches (including hairline cracks and pinhole leaks) in the fuel pin sections that will be exposed to an oxidizing atmosphere. This can be done by a review of records (for example, sipping records) or 100% eddy current inspection of assemblies;

3. Determine the time-at-temperature profile of the rods while they are exposed to an oxidizing atmosphere and calculate the expected oxidation to determine if a gross breach would occur.

Rather than terminating the use of air for blow-down, Constellation Energy proposes to restrict its use to only those canisters containing fuel which meets the requirements of option 2. Our current fuel loading procedures for the NUHOMS-32P require that fuel selected for loading have no known or suspected cladding failures. Reference documentation such as Reactor Coolant System chemistry or fuel sipping records is generally provided for each assembly selected to substantiate this determination. In recent years, we have conducted vacuum canister sipping campaigns, which represent the best available technology for detecting cladding breaches, specifically to qualify several hundred fuel assemblies as free of cladding failures and thus eligible for loading in the ISFSI. We propose that an inert gas would only be required for blow-down if the canister contained an assembly for which reference documentation was not available to substantiate the claim that the assembly did not contain rods with breached cladding. The response to RAI 8-2 below describes in more detail proposed changes to our ISFSI Technical Specification definitions to make them consistent with our current high standards for qualifying fuel selected for loading as free of cladding failures.

RAI 8-2:

Clarify the proposed contents of the package and provide separate definitions for intact and undamaged fuel in the TS.

The TS specify that the fuel shall be intact but can also include structural defects such as pinhole leaks.

These statements are not consistent with the guidance provided in ISG-l, Revision 2, "Damaged Fuel."

This information is needed to evaluate compliance with 10 CFR 72.122 5

ATTACHMENT (1)

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RAI 8-2 Response:

We propose to insert definitions for intact and undamaged fuel into the CCNPP ISFSI Technical Specifications, and revise Technical Specification 3.1.1 to use those definitions. See Attachment (2) for the marked up pages. The definition for undamaged fuel is based on the default definition of damaged fuel from American National Standards Institute Standard N14.33-2005, "Storage and Transport of Damaged Spent Nuclear Fuel," as provided on page 10 of ISG-1, which was then altered to allow credit for the performance based approach described in ISG-1 (page 5). The definition of intact fuel is also based on ISG-1, which indicates that intact fuel is a subset of undamaged fuel that is also known to have unbreached cladding. As discussed in the response to RAI 8-1 above, use of air for blow-down will be restricted to NUHOMS-32P DSCs containing intact fuel.

REFERENCES (1) Letter from Mr. J. A. Spina (CCNPP) to Document Control Desk (NRC), dated June 15, 2009, License Amendment Request: Allow Increased Burnup Fuel to be Loaded into NU-HOMS-32P Dry Shielded Canister (2) Letter from Mr. J. Goshen (NRC) to Mr. J. A. Spina (CCNPP), dated January 21, 2010, First Request for Additional Information for License Amendment Request No. 9 to Materials License No. SNM-2505, Calvert Cliffs Independent Spent Fuel Storage Installation (TAC No. L24350) 6

ENCLOSURE (I TRANSNUCLEAR CALCULATION 1095-57 Calvert Cliffs Nuclear Power Plant, LLC February 18, 2010

I,

• C#~-o63l L{

COM!TOLLED Copw.p A Form 3.1-1 TRANSNUCLEAR Calculation Approval Sheet Project Name: NUHOMS-32P Project #: 10950 Calculation

Title:

Thermal Analysis of Vacuum Drying Calculation #: 1095-57 Draft/Revision #: 0 DCR #: -

Number of pages: 5 Number of CDs attached: 1 If original Issue, 10CFR72.48 review required?

[X] No (explain) [ ] Yes, SR No. _

This calculation is performed in support of the licensee, CCNPP

1. This calculation is complete and ready for Independent review Originator's Signature Z ZJ e4 Date: ___-_

2. This calculation has been checked for consistency, completeness, and arithmetic correctness.

Checker Signature 1 /-' Date: 7/1/43

3. Calculation preparation and check complies with procedure - package Is complete PE's Signature Date: 81 _(

TRANSNUCLEAR, INC.

Tnt. Thermal Analysis of Vacuum Drying sKr 1 or 5 CAM N 1095-57

_RV. 0 1.0 Purpose To determine fuel cladding temperatures of the NUHOMS-32P during vacuum drying.

2.0 References

1) Not Used

2) Not Used,

3) Calculation 1095-29, Rev. 0, DSC Thermal Analysis - Normal Storage Conditions

4) Calculation 1095-38, Rev. 0, Effective Fuel Properties for Vacuum Drying 00&o4

5) CA03943, Rev. 0, ISFSI - Temperature and Heat Up of the Cask/DSC

6) ANSYS Computer Code and User's Manuals, Volumes 1-4, Rev. 6.0. See Test Reports E-19197 for validation of computer code.

7) ANSYS files: /Calc1095-40/ DSC.vd.db, DSC vd.rth, DSC_vd.mac

8) Calculation 1095-6, Rev. 0, Transfer Thermal Analysis, 103 OF Ambient CAW-"

9) Rohsenow et. al., Handbook of Heat Transfer Fundamentals, 2nd edition, 1985.

10)Bolz et al, Handbook of tables for Applied Engineering Science, 2 n edition, 1973.

11)Not Used 12)Calvert Cliffs Independent Spent Fuel Storage Installation, Volume 1, USAR, Rev.

11 13)Calculation 1095-53, Rev. 0, Transfer Analysis of Transfer Case with Poison Material 3.0 Assumptions and Discussion According to Reference 12 for the vacuum drying procedure, the cask cavity drain port is opened, and the water is drained from the annulus until the water level is approximately 12" below the top edge of the DSC shell. The DSC air space will be purged with filtered plant air, before the welding of the top shield plug begins. Engaging compressed helium or compressed air then removes the remaining water from the DSC cavity.

DSC Model The finite element model developed in Reference 3 was used to perform the vacuum drying thermal analysis. The temperature distributions of the fuel assemblies are determined under steady state conditions.

According to the description the of vacuum drying procedure, the DSC shell is in contact with the water during the entire procedure. The maximum basket temperature is expected near the axial center of the active fuel. Since, the water in the annulus does not produce any steam, the maximum accessible temperature of water is the saturation temperature.

TRANSNUCLEAR, INC.

Tm.e Thermal Analysis of Vacuum Drying sHEY 2 OF 5

_c__c_

NO 1095-57 ReV. 0 The following figure shows the location of the mid length of the active fuel during vacuum drying.

DSC 172.75" Active Fuel length, 136.7" Water Transfer ask 5.565" The saturation temperature of water at the mid length of the active fuel controls the temperature of the DSC shell. In order to find the saturation temperature of the water at that depth, first the water pressure at the mid length of the active fuel length is calculated.

P,= P +(p*g*h)*l P= Saturated water pressure, psia P Atmospheric

=tm pressure, 14.7 psi p = Water density at 212 F, 59.81 Ibm/ft3 [10]

g = gravitational acceleration constant, 32.174 ft/s 2 h = depth of water = 172.75 - (6.5 + 5.565 + 2136.7/2) - 12 = 80.335 in.

gc = conversion constant, 32.174 lbm*ft/lbf*s

• , 80.335in. *32.1744t)

• 1l P, = 14.7 psi+ ( 5 9 . 81I ft 32 17 4 Ibm *ft 3bf *S2 P,, = 17.48 lpsia using the steam tables from Ref. 10 gives the saturated temperature for water T,, = 220.7°F

TRANSNUCLEAR, INC.

rM Thermal Analysis of Vacuum Drying SHEE 3 O 5 cc. NO 1095-57 REV. 0 However, local boiling and the growth and collapse of steam bubbles cause a very large heat transfer coefficient that will tend to keep the canister shell temperature close to the water temperature. Evaporation from the surface of the annulus, together with the convection from the cask surface, tend to maintain the temperature constant.

Therefore, a constant temperature of 215 OF on the DSC shell is considered to be a reasonable assumption for this calculation.

All the material properties were identical to Reference 3 except for the back fill gas, fuel properties, and basket plates. Back fill gas properties are changed to those of air at 0.1 bar. The properties of the fuel assembly for vacuum conditions are calculated in Reference 4. Thermal properties Aluminum/Poison plates are taken from Reference

13. These values are listed below.

Air Conductivity at 0.1 bar Tv T K49J / k (mat 7, Effective Conductivity (mat 9, 20)**

(K) .. (F) (Vym=K)i (Btu/hr-in- F) (Btu/hr-in-°F) 300 80 70.0263-/7, 0.00127 0.002533 400 260 0. 0336 0.00162 0.003236 500 440 0.0403 0.00194 0.003882 600 620 0.0466 0.00224 0.004489 800 980 0.0577 0.00278 0.005558 1000 1340 0.0681 0.00328 0.00656

• • gaps noded as 2x size of design gaps Effective Conductivity Borated Aluminum/Al-1 100 Combination T Effective w/ Borated [131 (OF) (Btu/hr in OF) (Mat 23) 70 13.3 100 13.3 150 13.3 200 13.4 250 13.4 300 13.3 350 13.3 400 13.2 500 13.1 600 13.0 700 12.9 800 12.8

TRANSNUCLEAR, INC.

Thermal Analysis of Vacuum Drying SHee 4 OF 5 CAC. NO 1095-57 Effective A-i 100 Conductivity T Al-1100 [131

(°FI (Btu/hr in OF) (Mat 21) 70 14.8 100 14.7 150 14.5 200 14.3 250 14.2 300 14.0 350 13.9 400 13.8 Effective Fuel Conductivity Average Fuel Effective Radial Temperature Fuel Conductivity (OF) (Btu/hr-in-OF) 175.728 0.0080 258.740 0.0103 345.786 0.0133 435.945 0.0169 528.502 0.0213 622.858 0.0265 Since the conductivity of air is significantly lower than helium, radiation between the rails and the DSC across the gap contributes considerably to the heat transfer.

Radiation heat transfer between the DSC inner surface and the rail outer surfaces is modeled using one radiation super element matrix within /AUX12 processor. The radiation superelement includes inner surface of the DSC and the outer surfaces of the rail. SHELL57 elements are superimposed over the radiating surfaces for creation of the super-element. These elements are unselected prior to the solution of the finite element models.

Since, the water temperature in the annulus between the DSC and the transfer Cask is considered to be at 215 'F, and the ambient temperature will not exceed 103 OF. All the component temperatures of the transfer cask including the neutron absorbing resin are in a range between 215 OF and 103 OF. Therefore, the thermal limits for the transfer cask will not be exceeded.

3.1 Thermal Design Criteria

• A maximum fuel cladding temperature limit of 570 °C (1058 OF) is set for the fuel assemblies as concluded in Reference 8.

ME Thermal Analysis of Vacuum Drying SHET 5 OF 5 CA-C. NO 1095-57 REV. 0 4.0 Results and Conclusions The temperature distribution of the DSC cross section is shown in the figure below ANSYS 6.0 NODAL SOLUTION STHP=2 sue =1 TM=2 TENP Si2l1 =215 SXX =74g.965 215 274.441 333.881 393.32Z 452.762 512.203 571.643 631.084 690.52 4 749.965 Maximum Component Temperatures in NUHOMS-32P packaging for vacuum drying.

Component Maximum Temperature Thermal Limits

(_---) (CF) (OF)

Canister Outer Shell 218 Rails 502 Fuel Compartment 704 ...

Aluminum Basket Plates 700 ---

Stainless Steel Bars 703 ---

Fuel Cladding 750 1058 All components remain below thermal design criteria during vacuum drying procedure.

/r* ~~ .,. ~A * -*7

/ r_

ENCLOSURE (2)

TRANSNUCLEAR SPECIFICATION E-18851, REV. 7 Calvert Cliffs Nuclear Power Plant, LLC February 18, 2010

CONTROLLED COPY #: 1-01 UNCONTROLLED IF PRINTED FILE NUMBER#:

A TRANSNUCLEAR AN ARBVA COMPANY SPECIFICATION PAGE: 1 of 36 SPECIFICATION NO: E-18851 PROJECT NAME: NUHOMS@- 32P DSC PROJECT NO: 10950 CLIENT: Calvert Cliffs Nuclear Power Plant Project SPECIFICATION TITLE:

Design Criteria for the NUHOMS - 32P Storage System for Calvert Cliffs Nuclear Power Plant

SUMMARY

DESCRIPTION:

This document specifies the requirements for upgrading the design from a NUHOMS - 24P to NUHOMS - 32P Dry Storage System for Calvert Cliffs Nuclear Power Plant.

DCR PREPARER VERIFIER APPROVER QA APPROVAL NO. SIGNATURE/DATE SIGNATURE/DATE SIGNATURE/DATE SIGNATUREMDATE 11-11-o5 -- o W.S teln 6 10950-29 Jeff Gagne P. Shih 11/11/05 Jf Gaghe1/1/0 Jeff Ga7ne 11/11/05 Form 5.2-1, Revision 0

A TRANSNUCLEAR AN AREvA COMPANY SPECIFICATION NO: E-18851 REVISION: 6 PROJECT NO: 10950 PAGE: 2 of 36 TABLE OF CONTENTS PAGE 1.0 S C O P E................................................................................................................... 3 2.0 APPLICABLE DO CUM ENTS ............................................................................. 3 3.0 G ENERA L DESC R IPTIO N................................................................................ 8 4.0 D ESIG N R EQ UIR EM ENTS ................................................................................ 8 5.0 MATERIAL REQ UIREM ENTS .......................................................................... 17 6.0 QUALITY ASSURANCE REQUIREMENTS .................................................... 19 TABLE OF TABLES Table 1 PWR Fuel Assembly Design Characteristics ......................................... 20 Table 2

SUMMARY

OF NUHOMS - 32P SYSTEM DESIGN LOADINGS ....... 21 Table 3 HSM ULTIMATE STRENGTH REDUCTION FACTORS ........................ 27 Table 4 HSM LOAD COMBINATION METHODOLOGY ..................................... 28 Table 5 DSC DESIGN LOAD COMBINATIONS .................................................. 29 Table 6 TRANSFER CASK LOAD COMBINATION ............................ 30 Table 7 STRUCTURAL DESIGN CRITERIA FOR DSC ..................................... 31 Table 8 STRUCTURAL DESIGN CRITERIA FOR DSC SUPPORT ASSEMBLY ... 32 Table 9 STRUCTURAL DESIGN CRITERIA FOR ON-SITE TRANSFER CASK .... 33 Table 10 STRUCTURAL DESIGN CRITERIA FOR BOLTS ................................. 34 Table 11 Basket Stress Limits......................................... 35 Table 12 Therm al Load C ases ............................................................................. 36 Form 5.2-1, Revision 0

TRANSNUCLEAR AN AREvA COMPANY SPECIFICATION NO: E-18851 REVISION: 7 PROJECT NO: 10950 PAGE: 3 of 36 1.0 SCOPE This document specifies the requirements for upgrading the design from a NUHOMS-24P to NUHOMS-32P Dry Storage System for Calvert Cliffs Nuclear Power Plant. The design will be based on the NUHOMS design concept of horizontal storage, and is intended to be compatible with existing Horizontal Storage Module (HSM) and Transfer Cask system. General design requirements include structural, thermal, nuclear criticality safety and radiological protection criteria. The design and operation requirements for the Horizontal Storage Module (HSM) and the Transfer Cask system are specified in References 2.4.1 and 2.4.2.

2.0 APPLICABLE DOCUMENTS Unless otherwise noted in this specification, the documents, codes and standards referenced by this Specification shall be the revision, edition, addenda, and/or amendment in effect on October 7, 1988.

2.1 Codes and Standards 2.1.1 ASME Boiler and Pressure Vessel Code,Section III, Division 1, Subsections NB, NC, and NG, and Appendicesl1998, including 1999 addenda. (Transfer Cask, 1983) 2.1.2 ASME Boiler and Pressure Vessel Code,Section II, Materials Specifications, Parts A, B, C, and D, 1998, including 1999 addenda (Transfer Cask, 1992) 2.1.3 ASME Boiler and Pressure Vessel Code,Section V, 1998 including 1999 Addenda 2.1.4 ASME Boiler and Pressure Vessel Code,Section IX,Welding and Brazing Qualifications, 1998, including 1999 addenda 2.1.5 ANSI/ANS 57.9, "Design Criteria for an Independent Spent Fuel Storage Installation (Dry Storage Type)". 1992.

2.1.6 ANSI N14.5, "Leakage Tests on Packages for Shipment of Radioactive Materials". 1987.

Form 5.2-1, Revision 0

'A TRANSNUCLEAR AN AKEVIA COMPANY SPECIFICATION NO: E-18851 REVISION: .7 PROJECT NO: 10950 PAGE: 4 of 36 2.1.7 ANSI N14.6, "Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds or More," 1978 and June, 1993.(Transfer Cask, 1983) 2.1.8 ANSI N45.2, 1977, "Quality Assurance Program Requirements for Nuclear Power Plants" 2.1.9 ANSI N45.2.11, 1974, "Quality Assurance Program Requirements for Design of Nuclear Power Plants" 2.1.10 ANSIY14.5M-1982, "Dimensions and Tolerancing" 2.1.11 ANSI/ANS 57.2-1983, "Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants".

2.1.12 ANSI 8.17-1984, "Criticality Safety Criteria for Handling, Storage, and Transportation of LWR Fuel Outside Reactors".

2.1.13 American Nuclear Society, "American National Standard for Neutron and Gamma-Ray Flux to Dose rate Factors", ANSI/ANS 6.1.1-1977, LaGrange Park, Illinois.

2.1.14 ANSI N45.2.1 "Cleaning of Fluid Systems and Associated Components During Fabrication Phase of Nuclear Power Plants," 1980.

2.1.15 American National Standard, "Building Code Requirements for Minimum Design Loads in Buildings and Other structures". ANSI / 58.1-1982.

2.1.16 ASME Boiler and Pressure Vessel Code,Section VIII, Division 1,1983.

(Transfer Cask only) 2.1.17 AISC (Manual of Steel Construction) 8 th edition.

2.1.18 ASNT, SNT-TC-1A "Recommended Practice for Nondestructive Testing Personnel Qualification and Certification," 1992.

2.1.19 AWS D1.1 - 88, "Structural Welding Code - Steel" 2.1.20 AWS A2.4 - 86 "Weld Symbols" 2.1.21 SSPC - SP6, "Surface Preparation Specification No. 6 Commercial Blast Cleaning" Form 5.2-1, Revision 0

A TRANSNUCLEAR AN AREVA COMPANY SPECIFICATION NO: E-18851 REVISION: -7 PROJECT NO: 10950 PAGE: 5 of 36 2.1.22 ASTM A20 / A20M, "General Requirements for Steel Plates for Pressure Vessels" 2.1.23 ASTM A.380, "Standard Recommended Practice for Cleaning and Descaling Stainless Steel Parts, equipment and Systems" 2.1.24 ASTM A.480, "Standard Specification for General Requirements Flat-Rolled Stainless Steel and Heat-Resisting Steel Plate, Sheet and Strip" 2.1.25 ASTM A.484, "Standard Specification for General Requirements for Stainless and Heat-Resisting Wrought Steel Products (Except Wire)"

2.1.26 ASTM B29, "Standard Specification for Pig Lead" 2.2 Federal Regqulations 2.2.1 Title 10, Code of Federal Regulations, Part 20, "Standards for Protection Against Radiation."

2.2.2 Title 10, Code of Federal Regulations, Part 21, "Reporting of Defects and Noncompliance",

2.2.3 Title 10, Code of Federal Regulations, Part 50, Appendix B, "Quality Assurance Requirements for Nuclear Power Plants and Fuel Reprocessing Plants."

2.2.4 Code of Federal Regulations, Title 10, Part 71, Subpart H - Packaging and Transportation of Radioactive Materials, Quality Assurance.

2.2.5 Title 10, Code of Federal Regulations, Part 72, "Licensing Requirements for the Storage of Spent Fuel in an Independent Spent Fuel Storage Installation" 2.2.6 Title 49, Code of Federal Regulations, Part 173, "General Requirements for Shipments and Packaging" 2.2.7 Title 49, Code of Federal Regulations, Part 393, "Parts and Accessories" 2.2.8 USNRC, "Missiles Generated by Natural Phenomena," Standard Review Plan NUREG-0800 (1981)

Form 5.2-1, Revision 0

TRANSNUCLEAR AN AREVA COMPANY SPECIFICATION NO: E-18851 REVISION: 7 PROJECT NO: 10950 PAGE: 6 of 36 2.2.9 Title 10, Code of Federal Regulations, Part 50, "Domestic Licensing of Production and Utilization Facilities" 2.3 NRC Bulletins, Re-gulatory Guides, NUREG Documents, and EPA Federal Guidance Reports NOTE - NUREG documents are for guidance only, these documents do not impose requirements.

2.3.1 NRC Regulatory Guide 1.60, "Design Response Spectra for Seismic Design of Nuclear Power Plants.", Revision 1, 1973.

2.3.2 NRC Regulatory Guide 1.61, "Damping Values for Seismic Design of Nuclear Power Plants.", 1973 2.3.3 NRC Regulatory Guide 1.76, "Design Basis Tornado for Nuclear Power Plants" 2.3.4 NRC Regulatory Guide 1.92, "Combining Modal Responses and Spatial Components in Seismic Response Analysis" 2.3.5 NRC Regulatory Guide 3.54, "Spent Fuel Heat Generation in a Independent Spent Fuel Storage Installation".

2.3.6 NRC Regulatory Guide 3.60, "Design of an Independent Spent Fuel Storage Installation (Dry Storage)"

2.3.7 NUREG CR-5661, "Recommendations for Preparing the Criticality Safety Evaluation of Transportation Packages", 1997.

2.3.8 NUREG - 1536 Standard Review Plan for Dry Cask Storage System 2.3.9 NRC BULLETIN 96-04: Chemical, Galvanic, or Other Reactions In Spent Fuel Storage and Transportation Casks, July 5, 1996.

2.3.10 ISG-15, Rev 0, Materials Evaluation Form 5.2-1, Revision 0

TRANSNUCLEAR AN AREVA COMPANY SPECIFICATION NO: E-18851 REVISION: 7I PROJECT NO: 10950 PAGE: 7 of 36 2.3.11 ISG-1 1, Rev. 3, "Cladding Considerations for the Transportation and Storage of Spent Fuel" 2.3.12 ISG-2, Revision 0, "Fuel Retrievability" 2.4 Technical Reports and Documents 2.4.1 Not Used 2.4.2 Calvert Cliffs Independent Spent Fuel Storage Installation Updated Safety Analysis Report, Rev. 10.

2.4.3 NUTECH Report NUH -002, Topical Report for the NUTECH Horizontal Modular Storage System for Irradiated Fuel, NUHOMS - 24P,"

Revision 1A, July 1989.

2.4.4 Calvert Cliffs Independent Spent Fuel Storage Installation Technical Specification, Amendment 5.

2.4.5 Calvert Cliffs NUHOMS 32P DSC Design Specification SP-0564C, rev. 3 2.4.6 Calvert Cliffs Design Specification SP-0564, rev. 10 2.4.7 "Materials License SNM-2505 Technical Specifications," Calvert Cliffs Nuclear Power Plant ISFSI, latest addition.

2.4.8 Electric Power Research Institute Report NP-7419 Project 2813-9, "Fuel Assembly Behavior Under Dynamic Impact Loads De to Dry-Storage Cask Mishandling," Final Report, July 1991.

2.4.9 CCNPP Letter "DES Support for Increased Control Element Assembly (CEA) Weight," March 27, 2001; NEU 01-047.

2.5 QA Documents 2.5.1 Transnuclear Quality Assurance Program Form 5.2-1, Revision 0

A, TRANSNUCLEAR AN AREVA COMPANY SPECIFICATION NO: E-18851 REVISION: 7 PROJECT NO: 10950 PAGE: 8 of 36 3.0 GENERAL DESCRIPTION The NUHOMS-32P Dry Shielded Canister (DSC) shall be designed to provide storage of spent fuel in a Horizontal Modular Storage (HSM) system for 32 PWR fuel assemblies. The DSC shall also be compatible with the Calvert Cliffs Nuclear Power Plant (CCNPP) transfer cask. The NUHOMS-32P system is an upgrade of the NUHOMS-24P system licensed in accordance with the requirements of the Code of Federal Regulations, Title 10, Part 72 (10 CFR 72).

The NUHOMS-32P system shall accommodate 32 intact PWR fuel assemblies with fuel assembly characteristics defined in Table 1.

4.0 DESIGN REQUIREMENTS 4.1 General Desigqn Criteria The general requirements of the NUHOMS-32P system are listed below.

Specific component requirements are provided in subsequent sections.

NUHOMS-32P system shall maintain irradiated fuel assemblies in a subcritical state and provide confinement and shielding during handling, transfer, and storage. The NUHOMS-32P system consists of three principal components: Dry Shielded Canister (DSC), Horizontal Storage Module (HSM), and Transfer Cask (TC).

- The DSC shall be designed to stand vertically in the Transfer Cask. Lifting blocks, lifting rods and a lifting fixture and alignment marks shall be provided to properly orient the DSC with respect to the transfer cask. The lifting fixture maintains the round DSC shape and is designed to allow the DSC to be lifted in vertical orientation using four lifting points. The DSC shall be rotated from horizontal to vertical using a rotating fixture.

- The shell assembly of the DSC will be identical to the 24P design, except for changes to meet ASME Code requirements or leak testing improvements.

Leak test to at least 22.5 psig internal pressure. The maximum leak rate shall be less than 10-7 atm-cc/sec (no change in 24P shell assembly design requirements).

- The DSC shall provide a structural support and containment barrier for horizontal storage of 32 intact, PWR fuel assemblies in a dry, helium Form 5.2-1, Revision 0

TRANSNUCLEAR AN AREVA COMPANY SPECIFICATION NO: E-18851 REVISION: 7 PROJECT NO: 10950 PAGE: 9 of 36 atmosphere. DSC will be able to slide horizontally into and out of the transfer cask and onto the structural steel support assembly within the HSM.

- A basket assembly that locates and supports the fuel assemblies in the DSC, transfers the heat to the DSC outer surface and provides neutron absorption as necessary to satisfy nuclear criticality requirements.

- The HSM is a reinforced concrete structure, which provides physical and radiological protection of the DSC during storage operations.

- Heat from the DSC outer surface is transferred to the outside environment primarily by radiation and natural convection of air through the HSM and secondarily by natural convection and radiation from the concrete surface to the ambient.

- The TC shall provide physical and radiological protection during fuel handling and transfer operations from auxiliary building to the HSM.

- A gamma shield (lead) surrounds the DSC (transfer cask).

- A neutron shield surrounds the gamma shield, enclosed in an outer stainless steel shell that provides additional radiation shielding against neutrons (transfer cask).

- Sets of upper and lower trunnions provide support, lifting, and rotation capability for the transfer cask.

- The DSC design shall include provisions to positively align the DSC basket assembly with the DSC shell, top shield plug, and top cover plate.

4.1.1 Design Basis Fuel Characteristics The NUHOMS-32P system shall be capable of handling, transfer, and storage for PWR fuel assemblies with the characteristics included in Table 1.

4.1.2 Design Basis Pressures - Storage For storage considerations, it should be assumed that none of the fuel rods are failed for normal and off normal conditions. It is also assumed that all of the fuel rods will have failed following a design basis accident Form 5.2-1, Revision 0

TIRANSNUCILEAR AN.AREVA COMPANY

.SPECIFICATION NO: E-1 8851 REVISION: 7 PROJECT NO: 10950 PAGE: 10 of 36 event. A minimum of 100% of the fill gas and 30% of the fission gases (e.g., H-3, Kr and Xe) within the ruptured fuel rods should be assumed to be available for release into the DSC cavity.

4.1.3 Geometry and Weight Requirements To accommodate the NUHOMS Horizontal Storage Module and on-site transfer cask system, the canister shall be a right circular cylinder with a nominal outside diameter of 67.25 in. and 176.50 in. length, with a combined size, circularity and straightness tolerance of +/- 0.20 in. The maximum weight on the hook is 125 Tons.

4.2 NUHOMS - 32P Canister Structural Design Requirements Table 2 provides a summary description of the various loads requiring evaluation.

A summary of the load combination is included in Table 5.

4.2.1 NUHOMS -32P Canister Structure Design Criteria 4.2.1.1 NUHOMS -32P DSC Canister Shell Stress Limits

- The stress limits for the DSC canister shell are taken from the ASME Boiler and Pressure Vessel Code,Section III, Subsection NB, Article NB-3200 for normal condition loads (Level A) and Appendix F for accident condition loads (Level D).

- The stress due to each load shall be identified as to the type of stress induced, e.g. membrane, bending, etc.,

and the classification of stress, e.g. primary, secondary, etc.

- Stress limits for Level A and D service loading conditions are given in Table 7. Local yielding is permitted at the point of contact where the Level D load is applied. If elastic stress limits cannot be met, the plastic system analysis approach and acceptance criteria of Appendix F of Section III shall be used.

Reference to ASME, Section I1I, Subsection NB, Para.

NB-3223 and 3224 for Level B and Level C stress limits.

Form 5.2-1, Revision 0

A TRANSNUCLEAR AN ARE VA COMPANY.

SPECIFICATION NO: E-18851 REVISION: 7 '7 PROJECT NO: 10950 PAGE: 11 of 36 The allowable stress intensity value, Sm, as defined by the Code shall be taken at the temperature calculated for each service load condition. (see table 5) 4.2.1.2 NUHOMS -32P Canister Basket Stress Limits The basket fuel compartment wall thickness is established to meet heat transfer, nuclear criticality, and structural requirements. The basket structure must provide sufficient rigidity to maintain a subcritical configuration under the applied loads.

The primary stress analyses of the basket for Level A (Normal Service) and sustained Level D conditions do not take credit for the poison plates except for through thickness compression. The poison plate strength is, however, considered when determining secondary stresses in the stainless steel.

Normal Conditions The basis for the stainless steel fuel compartment section stress allowables is the ASME Code,Section III, Subsection NG. The primary membrane stress intensity and membrane plus bending stress intensities are limited to Sm (Sm is the code allowable stress intensity) and 1.5 Sm, respectively, at any location in the basket for Level A (Normal Service) load combinations. The average primary shear stress is limited to 0.6 Sm.

The ASME Code provides a basic 3Sm limit on primary plus secondary stress intensity for Level A conditions. That limit is specified to prevent ratcheting of a structure under cyclic loading and to provide controlled linear strain cycling in the structure so that a valid fatigue analysis can be performed.

Accident Conditions The basket shall be evaluated under Level D Service loadings in accordance with the Level D Service limits for components in Appendix F of Section III of the Code. The hypothetical impact accidents are evaluated as short duration Level D conditions. For elastic quasi-static analysis, the primary membrane stress (Pm) is limited to the smaller of 2.4Sm or 0.7Su and membrane plus bending stress Form 5.2-1, Revision 0

A TRANSNUCLEAR AN AREVA COMPANY

.SPECIFICATION NO: E-18851 REVISION: 7 PROJECT NO: 10950 PAGE: 12 of 36 intensities are limited to the smaller of 3.6Sm or 1.0Su. The average primary shear stress is limited to the smaller of 0.42 Su or 2(0.6Sm). When evaluating the results from the non-linear elastic-plastic analysis for the accident conditions, the general primary membrane stress intensity, Pm, shall not exceed 0.7Su and the maximum stress intensity at any location (PI or Pi + Pb) shall not exceed 0.9 Su.

The fuel compartment walls, when subjected to compressive loadings, are also evaluated to ensure that buckling will not occur. The critical loads for buckling of the basket should be calculated using ANSYS finite element Nonlinear Buckling Analysis. Dynamic analysis using LS-DYNA may also be used to calculate the buckling load. Reasonable safety factors for the allowable buckling loads should be provided to take into account material and geometrical imperfections.

Fusion Welds Fusion welds between the stainless steel plates and the stainless steel fuel compartments shall be qualified by testing. The minimum capacity shall be determined by shear test (pull test) of individual specimen made from production material.

The stress and load limits for the basket are summarized in Table 11.

4.3 On-Site Transfer Cask and HSM Design Requirements 4.3.1 On-Site Transfer Cask Structural Design Requirements The transfer cask is designed to the maximum practical extent as an ASME Class 2 component in accordance with the rules of the ASME Boiler and Pressure Vessel Code, Section I11,Subsection NC.

Table 2 provides a summary description of the various loads requiring evaluation. A summary of the load combination is included in Table 4.

The transfer cask design criteria are summarized in Tables 9 and 10.

4.3.2 Horizontal Storage Module Structural Design Requirements The NUHOMS reinforced concrete HSM is designed to meet the requirements CCNPP USAR.

Table 2 provides a summary description of the various loads requiring evaluation. A summary of the load combination and design criteria is Form 5.2-1, Revision 0

A TRANSNU.CLEAR AN AREVA COMPANY.

SPECIFICATION NO: E-18851 REVISION: 7 PROJECT NO: 10950 PAGE: 13 of 36 included in Table 4. The structural design criteria for DSC support assembly is included in Table 8.

4.4 Thermal Requirements Thermal properties of materials including material temperature limits are based on data from Calvert Cliffs ISFSI USAR (Reference 2.4.2). Following is a list of the thermal requirements obtained from Calvert Cliffs ISFSI USAR.

4.4.1 The peak cladding temperature of the fuel at the beginning of the long-term storage shall not exceed the NRC ISG-1 1 (Ref. 2.3.11) acceptance level of 4000 C (7520F).

4.4.2 Fuel cladding (zircaloy) temperature shall be maintained below 570 0 C (1058 0 F) for short-term accident conditions, short term off- normal conditions and fuel transfer operations (e.g. vacuum drying of the canister or dry transfer).

4.4.3 Total decay heat of 32 intact spent fuel assemblies is limited to 21.12 kW (0.66 kW per spent fuel assembly).

4.4.4 Normal storage (in the HSM) external ambient conditions, will use a lifetime average ambient temerature of 70'F and an average external insolation level of 82 Btu/hr-ft will be considered..

4.4.5 Normal transfer (in the transfer cask (TC)) will use an external ambient-'

temperature of -30F with no insolation, for winter extreme conditions and 103 0 F with an insolation level of 82 Btu/hr-ft 2 for summer extreme conditions.. Maximum temperatures of the DSC cladding and components will meet the requirements of ISG-11 (Ref. 2.3.11).

4.4.6 The off-normal storage and transfer will use ambient extremes based on maximum ambient temperatures of-3 0 F and 103 0 F and an insolation level of 127 Btu/hr-ft 2 .

4.4.7 Fuel Cladding and basket material temperatures should be calculated assuming steady state conditions during vacuum drying operations. If Form 5.2-1, Revision 0

A TRANSNUCLEAR AN AREVA COMPANY SPECIFICATION NO: E-18851 REVISION: 7 PROJECT NO: 10950 PAGE: 14 of 36 calculated temperatures are not acceptable, transient analysis should be performed assuming limited time period for vacuum drying operations.

4.4.8 HSM/DSC/Fuel materials shall be maintained within their minimum and maximum temperature criteria for normal, off-normal and accident conditions.

Thermal load cases are summarized in Table 12.

4.5 Shielding Requirements Predicted source terms and radiation dose rates shall be based on the transfer cask filled with the irradiated fuel types specified in Table 1. The neutron source term should be based on the minimum enrichment for the design basis burnup.

Form 5.2-1, Revision 0

TRANSNUCLEAR AN AREVA COMPANY SPECIFICATION NO: E-18851 REVISION: 7 PROJECT NO: 10950 1PAGE: 15 of 36 4.5.1 The transfer cask shall be designed to limit radiation exposure to both operators and the general public in accordance with ALARA.

4.5.2 For storage the radiation shielding must meet the requirements of 10CFR72.104 and 10CFR72.106.

4.5.3 After a design basis accident an individual at the boundary or outside the controlled area shall not receive a dose rate greater than 5 rem to the whole body or to any organ.

4.5.4 Doses calculated for workers and the public shall comply with the criteria in 10 CFR 20 and 72.

4.5.5 Gammas with energies from approximately 0.8 to 2.5 Mev will be considered as significant contributors to the dose rate.

4.5.6 The contribution from the irradiated fuel assembly hardware to the source term and the dose rate shall also be considered.

4.5.7 The flux-to-dose rate conversion factor shall be based on ANSI/ANS 6.1.1-1977.

4.5.8 Degradation of shielding material at higher temperature if applicable, shall be accounted for in the shielding evaluation.

4.5.9 The DSC shall be provided with adequate gamma shielding to maintain the maximum contact dose on the exterior surface of the Transfer Cask to 200 mRem/hr.

4.5.10 The DSC shall be provided with adequate gamma shielding to maintain the maximum contact dose on the exterior surface of the HSM shield door to 100 mRem/hr. The HSM sides and roof shall be limited to 20 mRem/hr.

4.6 Criticality Requirements 4.6.1 General Criticality Criteria 4.6.1.1 No credit for fuel burn-up shall be taken. However, credit for the soluble boron in the fuel pool shall be utilized ( Limited to <

2,450 ppm).

Form 5.2-1, Revision 0

TRANSNUCLEAR AN AREVA COMPANY SPECIFICATION NO: E-18851 REVISION: 7 PROJECT NO: 10950 PAGE: 16 of 36 4.6.1.2 No credit for burnable poison materials within the fuel assemblies as a neutron absorber shall be taken.

4.6.1.3 For a single cask or an array of casks, keff + 2a + bias and uncertainties shall not exceed an upper subcritical limit of 0.95 under all credible normal, off-normal, and accident conditions.

Model bias and benchmarking bias shall be accounted for in the criticality analysis. This methodology is equal to criticality requirements of keff shall not exceed 0.95 for all conditions with a 95% probability at a 95% confidence level including uncertainties. The requirements of ANSI/ANS 57.2 are included in ANSI/ANS 57.9 (Ref. 2.1.5). The upper subcritical limit (usl) is equal to the criticality acceptance criteria as designated in ANSI/ANS 57.2.

4.6.1.4 Only 75% of the poison material used in the basket assembly will be credited. Greater than 75% of the poison material can be credited ifthe requirements of NUREG CR-5661 (Ref.

2.3.7) are met.

4.6.1.5 Criticality control shall not require special loading patterns or special rotational orientation of the fuel'assemblies.

4.6.1.6 Consideration of full and optimum moderator density conditions over the 0.1 to 1.0 g/cc range during wet loading and unloading of fuel.

4.6.1.7 Misloading of at least two VAP (value added pellet) fuel assembly with enrichment of 5.00 to be evaluated due to the fact that borated moderator and the use of poison materials as required, are included for criticality control.

4.6.1.8 Effect of a collapsed fuel assembly after accident drop shall be evaluated.

4.6.2 Storage 4.6.2.1 In accordance with 10 CFR 71.124, the canister design (during all operational steps including handling, packaging, transfer and storage) shall prevent criticality during all normal, off-Form 5.2-1, Revision 0

A TRAINSNUCLEAR AN AREVA COMPANY SPECIFICATION NO: E-18851 REVISION: 7 PROJECT NO: 10950 PAGE: 17 of 36 normal, and accident conditions. Before a nuclear criticality accident is possible, at least two unlikely, independent and concurrent or sequential changes must occur.

4.6.2.2 The canister shall be designed and fabricated such that the spent fuel is maintained in a subcritical condition under all credible normal, off-normal, and accident conditions. (10 CFR 72.124(a) and 72.23(c)).

• The criticality analysis shall demonstrate that the fuel assembly used as the design basis is the most reactive.

" The criticality analysis must demonstrate that the cask remains subcritical for all credible conditions of moderation.

4.7 Confinement / Containment Criteria 4.7.1 Storage 4.7.1.1 The canister must maintain confinement of radioactive material within the limits of 10 CFR 72.236(l) and 10 CFR 20 under normal, off-normal, and credible accident conditions.

The canister must be designed and tested to meet the leak tight criteria defined in ANSI N14.5-1997 (Reference 2.1.9) 5.0 MATERIAL REQUIREMENTS 5.1 Specifications 5.1.1 Materials meeting the requirements of ASME B&PV Code,Section III, Article NB-2000, and the specification requirements of Section II,shall be used in the design to the maximum extent.

5.1.2 Detailed procurement specifications shall be required for other materials to assure that mechanical and other property values used in the design calculations will be met.

5.2 Properties Form 5.2-1, Revision 0

TRANSNUCLEAR AN AREVA COMPANY SPECIFICATION NO: E-18851 REVISION: 7 PROJECT NO: 10950 PAGE: 18 of 36 5.2.1 The material properties, stress intensity values and allowable stresses shall be obtained from the ASME B&PV Code,Section II, Part D.

5.2.2 For other materials, the source of material property data shall be identified and documented.

5.2.3 Materials shall be selected based on their corrosion resistance, susceptibility to stress corrosion cracking, embrittlement properties, and the environment in which they operate during normal and accident conditions.

5.3 Materials Suitability (Chemical, Galvanic and Other Reactions)

Materials suitability shall be reviewed in accordance with 10 CFR 72, NRC Bulletin 96-04 and 10CFR71.44 (d). Materials and construction shall be selected to assure that there will be no significant chemical, galvanic, or other reaction among packaging components and contents.

Materials shall be chosen that will preclude a galvanic effect which could lead to unacceptable fuel cladding corrosion or generate flammable gases in unacceptable quantities.

Material suitability evaluation should include:

7 the possible reaction from water in-leakage;

- the behavior of materials under irradiation; and

- the behavior of materials during all operations, e.g. operating temperatures and loading pool environment.

5.4 Protective Coatings The materials used for protective coatings (ifrequired) shall be compatible with the cask/canister materials, operating temperatures, loading pool environment and other interfacing materials or components. The exterior paint shall be easily decontaminated.

5.5 Emissivities Emissivity values for various surfaces important for heat transfer shall be specified in the calculations.

Form 5.2-1, Revision 0

TRANSNUCLEAR AN AREVA COMPANY SPECIFICATION NO: E-18851 REVISION: 7 PROJECT NO: 10950 PAGE: 19 of 36 Effects of Radiation Construction materials, including o-ring shall be compatible with the expected radiation levels.

5.8 Prohibited / Hazardous Materials The design shall not include sulfur, mercury, asbestos, low melting point metals, their alloys or components.

Materials in contact with pool water shall not release materials that contain chlorine or other halogens, sulfur, nitrates, mercury, lead, zinc, copper, tin, gallium, arsenic, antimony, bismuth, silver, cadmium or indium.

6.0 QUALITY ASSURANCE REQUIREMENTS The safety related components of the NUHOMS-32P Canister shall be designed, procured, fabricated and tested in accordance with the most recent revisions of Transnuclear's Quality Assurance Manual.

Form 5.2-1, Revision 0

A TRANSNUCLEAR AN AREvA COMPANY, SPECIFICATION NO: E-18851 REVISION: 7 PROJECT NO: 10950 *PAGE: 20 of 36 Table I PWR Fuel Assembly Design Characteristics Physical Parameters:

Fuel Design: 14x14 PWR by Westinghouse/ CE Cladding Material: Zircaloy 4 Fill Gas Helium Maximum Initial Fill Pressure (psia) (psig) 465 (50) ???

Maximum Assembly Weight 1450 lbs Number of Grid Spacers (including top and bottom fittings) 9 Radiological Parameters:

Maximum Burnup (Assembly Average) 52,000 MWd/MTUI31 Minimum Cooling Time As needed to reach .66KW(3)

Initial Fissile Content (Max. Initial Enrichment) 4.5 w/o U-235 Total Gamma Source per Assembly (max.) 1.63 x 1015 Mev/sec Total Neutron Source per Assembly (max.) 4.18 x 108 n/sec Max. Initial Uranium Content (MTU/assembly(2)) .400 kg/assembly Maximum Decay Heat 660 W/assembly Nominal Specific Power 32.2 MW / MTU Geometrical Parameters Maximum Assembly Length (incl. irradiation growth) (in.) 158 Max. Assembly Cross-section (in.) (2) 8.25 Fuel Density (% Theoretical) 95 Rod Pitch (in) 0.580 Number of Fueled Rods 176 Maximum Active Fuel Length (in) 136.7 Fuel Rod OD (in) 0.440 Clad Thickness (in) 0.028 Fuel Pellet OD (in) 0.3795 Number of Guide Tubes 5 Guide Tube OD (in) 1.115 Guide Tube Wall Thickness (in)(4) 0.04 "I Deleted.

(2) The open dimension of each fuel compartment cell is 8.5 in. x 8.5 in. minimum.

(3) Fuel assemblies which do not meet these requirements may be stored in the NUHOMS-32P system if the following conditions are met.

0 Neutron source per assembly must be less than or equal to 3.30x108 n/sec/assembly, with spectrum bounded by Table 3.1-4 of ISFSI USAR (Reference 2.4.2).

0 Gamma source per assembly must be less than or equal to 1.53x1 015 Mev/sec/assembly, with spectrum bounded by Table 3.1-4 of ISFSI SAR (Reference 2.4.7).

(4) Maximum assembly length (including irradiation growth) (in.) 158 Form 5.2-1, Revision 0

A TRANSNUCLEAR AN APREVA COMPANY SPECIFICATION NO: E-18851 REVISION: 6 PROJECT NO: 10950 PAGE: 21 of 36 Table 2

SUMMARY

OF NUHOMS - 32P SYSTEM DESIGN LOADINGS Component Design Load Type Design Parameters Applicable Codes ACI 349-85 and ACI 349R-85 o

Max. wind pressure: 397 psf NRC Reg. Guide 1.76 Design Basis Tornado Max. widpresued  : 397 ph and Max. Speed: 360 mph ANSI A58.1 1982 Max. Speed: 126 mph DBT Missile Type: Automobile 3967 lb., NUREG - 0800 8 in. diam. Shell 276 lb., Section 3.5.1.4 1 in. solid sphere Flood None Required ISFSI USAR Seismic Hor. acceleration: 0.15g ISFSI USAR Horizontal Vert. acceleration: 0.10g Storage Snow and Ice Maximum load: 200 psf ISFSI USAR Module (included in live loads) ANSI 57.9-1984 Dead Loads Dead weight including loaded ANSI 57.9- 1984 DSC (concrete density of 150 pcf)

Normal and off-normal DSC with spent fuel rejecting 21.12 kw of ANSI 57.9 - 1984 Operating Temperatures decay heat. Ambient air temperature range of -30F to 103 0 F Accident Condition Same as off-normal conditions With HSM vents ANSI 57.9 - 1984 Temperatures blocked for 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> or less Normal Handling Loads Hydraulic ram load equal to 25% of loaded ANSI 57.9- 1984 DSC weight: 23,750 lb. Enveloping Hydraulic ram load equal to 100% of loaded Off-normal Handling Loads DSC weight: 95, 000 lb. Enveloping ANSI 57.9- 1984 Form 5.2-1, Revision 0

A TRANSNUCLEAR AN AREVA tOMPANY SPECIFICATION NO: E-18851 REVISION: 6 PROJECT NO: 10950 PAGE: 22 of 36 Table 2

SUMMARY

OF NUHOMS - 32P SYSTEM DESIGN LOADINGS (Continued)

Component Design Load Type Design Parameters Applicable Codes Live Loads Design load: 200 psf ANSI 57.9 - 1984 Horizontal (includes snow and ice loads)

Storage Fire 1 Hr. forest fire 65 ft from HSM None Module Probability of liquidfied natural gas spill affecting HSM ISFSI USAR Explosions < 10- NUREG -0800 Tornado Wind and Tornado The DSC is protected by the Transfer Cask and HSM, ASME Code,Section III, Missile Loads therefore, not needed Component Flood Maximum water height: 50 ft. None Required 10CFR 72.72 Seismic Horizontal acceleration:1.0g1.5g Vertical acceleration: 3% critical damping ISFSI USAR Dead Loads Weight of loaded DSC: 91,133 lb. ANSI 57.9 - 1984 nominal, 95,000 lb. enveloping Normal Temperature DSC with spent fuel rejecting 21.12 kw decay heat.

Dry Ambient air temp = -30F to 103 0 F, insolation = 82 2 ANSI 57.9 - 1984 Shielded BTU/hr-ft 2, Off Normal insolation - 127 BTU/hr-ft Canister Internal pressure-Normal 15psig, Off Normal-50psig Normal Pressure Blowdown pressure-15psig ANSI 57.9 - 1984 Vacuum Press < 3torr, for minimum 30 min.

Off-Normal Hydrostatic Hydrostatic pressure of annulus water on the DSC ISFSI USAR Pressure, Water filled Cask plus atmospheric pressure (14.7 psi)

Normal Handling Loads Hydraulic ram load equal to 25 % of loaded DSC ANSI 57.9 - 1984 weight: 23,750 lb. enveloping Hydraulic Ram load equal to 100 % of loaded DSC ANSI - 57.9-1984 Off-normal Handling Loads weight: 95,000 lb. enveloping Form 5.2-1, Revision 0 f

TRANSNUCLEAR AN AREvA COMPANY SPECIFICATION NO: E-1 8851 REVISION: 6 PROJECT NO: 10950( PAGE: 23 of 36 Table 2

SUMMARY

OF NUHOMS -32P SYSTEM DESIGN LOADINGS (Continued)

Component Design Load Type Design Parameters Applicable Codes Equivalent static deceleration of 75g for vertical Accident Drop end drop and horizontal side drops, and 25g ISFSI USAR Dry corner drop with slapdown Shielded DSC internal pressure Of 100 psig -based on Canister Accident Internal Pressure 100% fuel cladding rupture with fill gas release, ISFSI USAR 10CFR and 30% fission gas release at an ambient air 72.122 (b) temperature of 103 0 F AISC Code for Structural Steel Dead Weight Loaded DSC plus self weight ANSI 57.9 - 1984 Dry Horizontal acceleration: 0.61 g Shielded Seismic Vertical acceleration: 0.39g ISFSI USAR Canister With 7% critical damping Support DSC reaction loads with hydraulic ram load Assembly Normal Handling Loads equal to 25% of loaded DSC weight: 24,000 lb. ANSI -57.9-1984 enveloping DSC reaction loads with hydraulic ram load Off-normal Handling Loads equal to -100 % of loaded weight: ANSI - 57.9-1984 95,000 lb. enveloping Form 5.2-1, Revision 0

A TRANSNUCLEAR AN AREVA COMPANY SPECIFICATION NO: E-18851 REVISION: 6 PROJECT NO: 10950 PAGE: 24 of 36 Table 2

SUMMARY

OF NUHOMS - 32P SYSTEM DESIGN LOADINGS (Continued)

Component Design Load Type Design Parameters Applicable Codes Automobile 3,967 lbs. NUREG-0800 8" diameter shell, 276 lbs. Section 3.5.1.4 Design Basis Max. wind pressure: 397 psf NRC Reg. Guide Tornado Wind Max. wind speed: 360 mph 1.76 and

___________________ANSI 58.1 -1982 Not included in design due to infrequent short Flood duration; use of cask restricted by ISFSI USAR 10CFR 72.122 (b)

Administrative controls Hor. ground accel.: 0.25g ISFSI USAR Vert. ground accel.: 0.17g On-site Transfer External surface temp. and circular section will Cask Snow and Ice preclude build-up of snow and ice loads when ISFSI USAR 10CFR 72.122 (b) cask is in use Vertical orientation, self weight with loaded DSC and water in cavity: 220,000 lbs. ANSI 57.9-1984 Enveloping Dead Weight Horizontal orientation self Weight with loaded DSC on Transfer skid: 214,494 lbs. Nominal, 215,000 lbs. enveloping ANSI 57.9-1984 Loaded DSC rejection 21.12 kw NOpralind Off-nrmal Decay heat. Ambient air ANSI 57.9 - 1984 Operating Temperatures Temperature range: -30 F to 103 0F Form 5.2-1, Revision 0

TRANSNUCLEAR AN AREVA COMPANY SPECIFICATION NO: E-18851 _*REVISION: 6 PROJECT NO: 10950* PAGE: 25 of 36 Table 2

SUMMARY

OF NUHOMS - 32P SYSTEM DESIGN LOADINGS (Continued)

Component Design Load Type Design Parameters Applicable Codes Upper lifting trunnions in fuel building:

Stress due to 6 x load < yield stress Stresses yield with Stress due to 10 x load < Ultimate strength and ultimate with Upper lifting trunnions on -site transfer ISFSI USAR Normal Handling Loads Lower support trunnions: Weight of loaded ASME Section III On-site cask during down loading and proportional ASME Section III Transfer weight of loaded cask during transit to HSM ANSI 57.9 -1984 Cask Hydraulic ram load / friction of moving DSC equal to 25% of DSC loaded weight:

24,000 lb. enveloping Bolts - service level A, B, and C Hydraulic ram load/jammed DSC equal to Off-normal Handling Loads 100% of DSC loaded ANSI 57.9-1984 Weight: 95,000 lb. enveloping Form 5.2-1, Revision 0

TRANSNUCLEAR AN AREVA COMPAN SPECIFICATION NO: E-18851 REVISION: 6 PROJECT NO: 10950 PAGE: 26 of 36 Table 2

SUMMARY

OF NUHOMS -32P SYSTEM DESIGN LOADINGS (Concluded)

Component Design Load Type Design Parameters Applicable Codes Equivalent static deceleration of 75g for vertical ISFSI USAR 10CFR Accident Drop Loads end drops And horizontal side drops, and 25g 72.122 (b)

On-site for corner drop and slapdown Transfer Cask Fire and Explosions None required ISFSI USAR ISFSI USAR 10CFR Internal Pressure N/A - DSC provides pressure boundary 72.122 (b)

Form5.2-1, Revision 0

A TRANSNUCLEAR AN AREVA COMPANY SPECIFICATION NO: E-18851 JREVISION: 6 PROJECT NO: 10950I PAGE: 27 of 36 Table 3 HSM ULTIMATE STRENGTH REDUCTION FACTORS Type of Stress Reduction Factor Flexure 0.9 Axial Tension 0.9 Axial Compression 0.7 Shear 0.85 Torsion 0.85 Bearing 0.7 Form 5.2-1, Revision 0

TRANSNUCLEAR AN AREVA COMPANY SPECIFICATION NO: E-18851 REVISION: 6 PROJECT NO: 10950- PAGE: 28 of 36 Table 4 HSM LOAD COMBINATION METHODOLOGY Case Load Combination Loading Notation No.

1 U = 1.4D + 1.7L U = Required strength of a cross section or member to resist design 2 U = 1.4D + 1.7L + 1.7H loads or their internal moments and force 3 U = 0.75 (1,4D + 1,7L + 1.7H + 1,7T +

1.7W) D = Dead Weight E = Earthquake Load 4 U = 0.75 (1.4D + 1.7L + 1.7H + 1.7T)

W= Wind Load 5 U=D+L+H+T+E F = Flood Induced Loads 6 U=D+L+H+T+F H = Lateral Soil Pressure Load 7 U=D+L+H+Ta L = Normal Condition Live Load T = Normal Condition Thermal Load Ta = Off-normal or Accident Condition Thermal Load Notes:

1. The HSM Load Combinations are in accordance with ANSI - 57.9. In case 6 flood loads (F) are substituted for drop loads (A)which are not applicable to the HSM.

2. The effects of creep and shrinkage are include in the dead weight load for cases 3 through 7.

3. Wind Loads are conservatively taken as Design Basis Tornado (DBT) loads. These include wind pressure, differential pressure, and missile loads. Case 3 was first satisfied without the tornado missile load. Missile loads were analyzed for local damage, over all damage, overturning and sliding effects Form 5.2-1, Revision 0

A TRANSNUCLEAR AN AREVA COMPANY SPECIFICATION NO: E-18851 REVISION: 6 PROJECT NO: 10950 PAGE: 29 of 36 Table 5 DSC DESIGN LOAD COMBINATIONS Accident Conditions Normal Operating Off-Normal Emergency Conditions Load Combinations Case Conditions Conditions 1 2 3 4 1 2 3 4 1 2 3 4 5 6 1 2 3 4 5 Vertical, DSC Empty X Dead--

Weight Vertical, DSC w/Water X

Horizontal, DSC w/Fuel X X X X X X X X X X X *X X X X 0

Inside HSM:70 (amb.) Normal X X X 0

Inside Cask:70 (amb.) Normal X X X X Inside HSM:103 0 F(amb.) Off-N X X Thermal __

Inside Cask:103 0 (amb.) Off-N X X X X X X Inside HSM: Accident (vent block) X Normal Operating Pressure X X X X X X X Hydrostatic X -

-j Internal Off-Normal X X X W Pressure -

Accident (inner boundary) X X X Accident (outer boundary) X X Handling Normal DSC Transfer X X Loads Off Normal Jammed DSC Loads X X X X X Acci. Cask Drop (End, Side or Comer Drop) X X Accident Seismic X ASME Code Service Level A A A A B B B B C C C C D D D D D Load Combination No A1 A2 A3 A4 B1 B2 B3 B4 C1 C3 C5 C6 D1 D2 D3 D4 D5 Form 5.2-1, Revision 0

A TRANSNUCLEAR AN AREVA COMPANY SPECIFICATION NO: E-18851 J.REVISION: 6 PROJECT NO: 10950I PAGE: 30 of 36 Table 6 TRANSFER CASK LOAD COMBINATION Normal Operating Off-Normal Accident Load Case Conditions Conditions Conditions 1 2 3 4 5 1 2 1 2 3 4 5 6 7 Dead Load/Live Load X X X X X X X X X X X X X X Thermal 70°F Ambient(1 ) X X X X X X X X X X w/DSC 103 0F Ambient X X X Handling Vertical X LoadsLasTilted X

(Critical Lifts) Horizontal X Handling Transport X X X Loads (Non- Critical DSC Transfer X X X Seismic X X Tornado Wind X Tornado Missile X Vertical x Drop Corner x Horizontal x ASME Code Service Level A A A A A B B C C D D D D D (1) The thermal stress distribution used in the normal condition load combinations is computed using the 103° ambient, off-normal, temperature distribution.

. ,,,..Form 5.2-1, Revision 0

TRANSNUCLEAR AN AREVA COMPANY SPECIFICATION NO: E-18851 REVISION: 6 PROJECT NO: 10950 PAGE: 31 of 36 Table 7 STRUCTURAL DESIGN CRITERIA FOR DSC Stress Value Item Stress Service Service Level D Type - Levels Service Level C Elastic A&B Analysis Plastic Analysis General Sm Greater of 1.2Sm or Smaller of 2 4 Sm Greater of 0.7Su or Sy Membrane S, or 0.7Su + 1/3(S, - Sj)

Lcall(2) Greater of 1.8Sm or 150% of Pmn DSC Membrane 1.5Sm 1.5Sy Limit(5) 0.9S" Primary + 3.0Sm N/A N/A N/A Secondary DSC Fillet Greater of and Partial Primary 0.55m 0.65Sm(6) or 0.50S, Smaller of 1.2Sm or 0.35Su(1°)

Penetration Primary + Smaller of 0.9Sm or Welds(3) Secondary 0.75Sm 0.75S, N/A DSC Closure 0'8Smn8) Greater of 0.8x 2 4 Sm (8)(11) 2 l0m or 0.8xSy(8) Smaller of 0.8 x 0.7Su or 0.8x Welds(7) Primary 0.7Sm9) Greater of 0.7x Smaller of 0.7 x 0.7Su or 0.7x 2.4Sm (9)(12) 1.2Sm or 0.7xS,(9) 0.8 x Primary + 3Sm(8) N/A N/A Secondary 0.7 x 3Sm(9)

Notes:

(I) Not Used (2) Includes full penetration welds.

(3) An efficiency factor of 0.5 has been applied for nonvolumetric inspected welds based on ASME Section VIII, Div. 1, and Table UW-1 2 No. 5.

.7.

(4) Local primary membrane stress, PL, shall not exceed 150% of the P, limit.

(5) For elastic analysis, an alternative limit for PL + PB is that the static or equivalent static loads shall not exceed 90% of the limit analysis collapse load using a yield stress which is the lesser of2.3S= and 0.3S., or 100% ofthe plastic analysis or test collapse load; for plastic analysis, an alternative to the primary stress intensity limits is that the static or equivalent static loads shall not exceed 90% of the limit analysis collapse load using a yield stress which is the lesser of 2.3Sm and 0.7S,, or 100% of the plastic analysis or test collapse load.

(6) For 0.5 efficiency factor, 0.6S., is used for the allowable stress, However, it should be noted that even though an efficiency factor of 0.5 is applied for all nonvolumetric inspected welds, an efficiency factor high than 0.5 is allowed for any individual welds installed by different methods. 0.65S=is allowed for welds at Service Level C.

(7) Criteria for closure welds are taken from Code Case N-595-2 and ISG-15:

(8)

For inner cover plate weld (weld between lead plug top casing plate and shell).

(9) For outer cover welds (weld between top outer cover plate and shell).

(10) 3 For plastic analysis, the criteria is greater of O.35S. or 0.5[Sy+ 1/ (S,-Sy)].

(Io) 3 For plastic analysis, the criteria is greater ofO.8 x 0.7S. or 0.8[Sy+l/ (Su-Sy)].

((2)

For plastic analysis, the criteria is greater of 0,7 x 0.7S, or 0.7[Sy+ [/3(Su-Sy)].

Form 5.2-1, Revision 0

TRANSNUCLEAR AN AREVA COMPANY SPECIFICATION NO: E-18851 FREVISION: 6 PROJECT NO: 10950 PAGE: 32 of 36 Table 8 STRUCTURAL DESIGN CRITERIA FOR DSC SUPPORT ASSEMBLY Stress Type Stress Values Tensile 0.60 Sy Compressive (See Note 1)

Bending 0.60 Sy (2)

Shear 0.40 Sy Interaction (See Note 3)

Notes:

1. Equation 1.5-1, 1.5-2 or 1.5-3 of the AISC Code (3.45) are used as appropriate.

2. If the requirements of Paragraph 1.5.1.4.1 (AISC Code) are met, an allowable bending stress of 0.66 Sy is used.

3. Interaction equations per the AISC Code are used as appropriate.

Form 5.2-1, Revision 0

A TRANSNUCLEAR AN AREVA COMPANY SPECIFICATION NO: E-18851 REVISION: 6 PROJECT NO: 10950 PAGE: 33 of 36 Table 9 STRUCTURAL DESIGN CRITERIA FOR ON-SITE TRANSFER CASK Stress Values Item Stress Type Service Levels A& B Service Level C Service Level D Primary Sm 1.2 Sm or Sy Smaller of 2.4 Sm Transfer Membrane or 0.7 S u Cask Primary 1.5 Sn 1.08 Sm Smaller of 3.6 4 Sm Structural Membrane + Bending or S u Shell Primary + Secondary 3.0 Sm N/A N/A Membrane and Smaller of N/A N/A Membrane + Bending Sy / 6 or Su/10 Trunnions (1)

Shear Smaller of 0.6 Sy /6 N/A N/A or 0.6 Su / 10 DSC Primary (per ISG-15) (per ISG-15) (per ISG-15)

Fillet Welds (2) Primary + Secondary (per ISG-15) (per ISG-15) N/A Notes:

1. These allowables apply to the upper lifting trunnions for critical lifts governed by ANSI N14.6. The lower support trunnions and the upper lifting trunnions for all remaining loads are governed by the same ASME Code criteria applied to the cask structural shell.

2. The weld efficiency factor should be per ISG -15 [2.3.10].

Form 5.2-1, Revision 0

TRANSNUCLEARk.

AN AREVA COMPANY SPECIFICATION NO: E-18851 REVISION: 6 PROJECT NO: 10950 PAGE: 34 of 36 Table 10 STRUCTURAL DESIGN CRITERIA FOR BOLTS Service Levels A, B, and C Average Service Stress < 2 Sm Maximum Service Stress < 3Sm Service Level D Average Tension Smaller of Sy or 0.7 Su Tension + Bending Su Shear Smaller of 0.6 Sy or 0.42 Su Interaction Interaction equation of Appendix F (F-1335.3) of ASME Code (3.14)

Form 5.2-1, Revision 0

TRANSNUCLEAR AN AREVA COMPANY SPECIFICATION NO: E-18851 _[REVISION: 6 PROJECT NO: 10950- PAGE: 35 of 36 Table 11 Basket Stress Limits Allowable Stresses Stress Category Normal Conditions(l) Accident Conditions(2)

Primary Membrane Lesser of General Pm Sm 2.4 Sm or 0.7 u(3)

Lesser of Local PL 1.5 Sm 3.6 Sm or 1.0 SU (3)

Primary Membrane + Bending Lesserof (Pm or PL) + Pb 1.5 Sm 3.6 Sm or 1.0 So(3)

Range of Primary + Secondary 3.0 Sm 2Sa for 10 cycies (4)

(Pm or PL) + Pb + Q 3.0 _______for_10_cycles Bearing Stress S, Not applicable Average. Primary Shear Stress 0.6 Sm Lesser of 0.42 Su, or 2(0.6Sm)

Compressive Stress limit Buckling(7) per See Section 4.2 NF-3322.1(c)

Fatigue Cumulative fatigue usage Not applicable Fatigue_ factor < 1 I I Notes:

1. ASME Code,Section III, Appendix NG, service level A

2. ASME Code,Section III, Appendix F, service level D

3. When evaluating the results from the nonlinear elastic-plastic analysis for the accident conditions, the general primary membrane stress intensity, Pm, shall not exceed the greater of 0.7Su or Sy+l/3(Su-Sy) and the maximum primary stress intensity at any location (PL or PL + Pb) shall not exceed 0.9 Su.

4. ASME Code Section I11,Appendix land Reg. Guide 7.6.

5. ASME Code, Section Ill, Appendix F, Para. F-1341.3

6. Reference to Section III, Subsection NB, Para. NB-3223 and NB-3224 for Level B and Level C stress limits.

7. Other acceptable criteria are also provided in Section III of the ASME Code and NUREG/CR-6322.

Form 5.2-1, Revision 0

TRANSNUCLEAR ANARMA COMPAN SPECIFICATION NO: E-18851 [REVISION: 6 PROJECT NO: 10950[ PAGE: 36 of 36 Table 12 Thermal Load Cases For DSC within HSM Load Ambient Temp. Module Position Wall Thickness Loaded DSC in Vents Insolance Case Conditions (OF) (ft) Adjacent HSMs Status (Btu/hr-ft2 )

1 Normal Storage 70 Interior 2 Yes Open 82 2 70 Interior 2 No Open 82 3 70 End Module 3 Yes Open 82 4 Off Normal Storage 103 Interior 2 Yes Open 82 5 -3 Interior 2 No Open 0 6 103 Interior 2 Yes Open 127 7 103 Interior 2 No Open 127 8 Accident 103 Interior 2 No Blocked 127 9 103 End Module 3 Yes Blocked 127 10 103 Interior 2 Yes Blocked 127 For DSC within Transfer Cask Load Conditions Ambient Temp. Insolance Case (OF) (Btu/hr-ft2 )

-> 11 Transfer Operation 103 127 Form 5.2-1, Revision 0

ATTACHMENT (2)

MARKED UP TECHNICAL SPECIFICATION PAGES Calvert Cliffs Nuclear Power Plant, LLC February 18, 2010

1.0 DEFINITIONS The following definitions apply for the purpose of these Technical Specifications:

a. ADMINISTRATIVE CONTROLS: Provisions relating to organization operating, emergency, and management procedures, recordkeeping, review and audit, and reporting necessary to ensure that the operations involved in the movement, transfer and storage of spent fuel at the Calvert Cliffs ISFSI are performed in a safe manner.

b. DESIGN FEATURES: Features of the facility associated with the basic design such as materials of construction, geometric arrangements, dimensions, etc., which, if altered or modified, could have a significant effect on safety.

c. FUEL ASSEMBLY: The unit of nuclear fuel in the form that is charged or discharged from the core of a light-water reactor (LWR). Normally, will consist of a rectangular arrangement of fuel and non-fuel held together by end fittings, spacers, and guide tubes.

d. FUNCTIONAL AND OPERATING LIMITS: Limits on fuel handling and storage conditions necessary to protect the integrity of the stored fuel, to protect employees against occupational exposures, and to guard against the uncontrolled release of radioactive materials.

e. LIMITING CONDITIONS: The minimum or maximum functional capabilities or performance levels of equipment required for safe operation of the facility.

f. LOADING OPERATIONS: Loading Operations include all cask preparation steps prior to cask transport from the auxiliary building area.

g. SURVEILLANCE INTERVAL: A surveillance interval is the interval between a surveillance check, test or calibration. Unless specifically stated otherwise, each surveillance requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval.

h. SURVEILLANCE REQUIREMENTS: Surveillance requirements include: (i) inspection, test, and calibration activities to ensure that the necessary integrity of required systems, components, and the spent fuel in storage is maintained; (ii) confirmation that operation of the installation is within the required functional and operating limits; and (iii) a confirmation that the limiting conditions required for safe storage are met.

0jicse-Page 1 of 13 Amendment 7

INSERT A INTACT FUEL ASSEMBLIES:

Fuel assemblies meeting the following conditions are considered intact fuel assemblies for the purpose of storage in the Calvert Cliffs ISFSI:

1) the assembly is undamaged, and

2) no known cladding breaches, as indicated by reactor operating records or fuel qualification testing (e.g., vacuum canister sipping, etc.).

INSERT B UNDAMAGED FUEL ASSEMBLIES:

Fuel assemblies meeting the following conditions are considered undamaged for the purpose of storage in the Calvert Cliffs ISFSI:

1) no deformation of the fuel rods such that structural, criticality safety or radiological design functions are adversely impacted (e.g., deformation other than uniform rod bowing that does not significantly open up the lattice spacing);

2) no missing fuel rods such that structural, criticality safety or radiological design functions are adversely impacted (e.g., a dummy rod that displaces a volume equal to, or greater than, the original fuel rod, placed in the missing rod location is acceptable);

3) no missing, displaced, or damaged structural components such that radiological and/or criticality safety is adversely affected (e.g., significantly changed rod pitch);

4) no missing, displaced, or damaged structural components such that the assembly cannot be handled by normal means (e.g. no consolidated fuel),

5) no gross cladding breaches (other than pinhole leaks or hairline cracks), as indicated by reactor operating records (or other records); and

6) no debris that would adversely impact the structural, criticality safety, or radiological design function.

3/4.1 FUEL TO BE STORED AT ISFSI LIMITING CONDITION FOR OPERATION 3.1.1 The spent nuclear fuel to be received and stored at the Calvert Cliffs ISFSI shall meet the following requirements:

(1) Only fuel irradiated at the Calvert Cliffs Units 1 or 2 may be used. (14 x 14 CE type PWR Fuel)

(2) Maximum initial enrichment shall not exceed 4.5 weight percent U-235.

(3) Maximum assembly a b riup shall ot exceed 7,000 megawatt-da s er metric ton uranium (Nut,.4 S - Zo) or- Sr,, R:70 ClLt+cjs F.,r (4) Minimum burnup shall excee .3-1.

(Applicable only to NUHOMS-24P.)

(5) Maximum heat generation rate shall not exceed 0.66 kilowatt per fuel assembly.

(6) Fuel shall have cooled as specified in ISFSI SAR Table 9.4.1.

(7) Maximum assembly mass including control components shall not exceed 1450 lb.(658 kg).

(8) Fuel shall be* -,, fýC

\ *'*

(9) F&asmWlskA.1 0 F APPLICABILITY: This specification is applicable to all spent fuel to be stored in Calvert Cliffs ISFSI.\

ACTION: If any fuel does not specifically meet the requirements for maximum burnup and post irradiation time (items 3 & 6 above), confirm to see if the requirements of Section 2.1 are satisfied. If any other requirements of the above specification are not satisfied, do not load the fuel assembly into a DSC for storage.

Fue- ~ko- tz,

& k .A- (N 0 4oM-.S- 32~

tA-k.

G~

Page 5 of 13 Amendment 7